10 CFR 51.107 - Public hearings in proceedings for issuance of combined licenses; limited work authorizations.

Code of Federal Regulations, 2011 CFR

2011-01-01

... the issuance of a combined license for a nuclear power reactor under part 52 of this chapter, the... or Director, Office of Nuclear Reactor Regulation, as appropriate. (b) If a combined license... authorization should be issued as proposed by the Director of New Reactors or the Director of Nuclear Reactor...

10 CFR 51.107 - Public hearings in proceedings for issuance of combined licenses; limited work authorizations.

Code of Federal Regulations, 2013 CFR

2013-01-01

... the issuance of a combined license for a nuclear power reactor under part 52 of this chapter, the... or Director, Office of Nuclear Reactor Regulation, as appropriate. (b) If a combined license... authorization should be issued as proposed by the Director of New Reactors or the Director of Nuclear Reactor...

Proceedings of a Symposium on Advanced Compact Reactor Systems

NASA Technical Reports Server (NTRS)

1983-01-01

Reactor system technologies suitable for a variety of aerospace and terrestrial applications are considered. Technologies, safety and regulatory considerations, potential applications, and research and development opportunities are covered.

Space nuclear power systems; Proceedings of the 8th Symposium, Albuquerque, NM, Jan. 6-10, 1991. Pts. 1-3

NASA Technical Reports Server (NTRS)

El-Genk, Mohamed S. (Editor); Hoover, Mark D. (Editor)

1991-01-01

The present conference discusses NASA mission planning for space nuclear power, lunar mission design based on nuclear thermal rockets, inertial-electrostatic confinement fusion for space power, nuclear risk analysis of the Ulysses mission, the role of the interface in refractory metal alloy composites, an advanced thermionic reactor systems design code, and space high power nuclear-pumped lasers. Also discussed are exploration mission enhancements with power-beaming, power requirement estimates for a nuclear-powered manned Mars rover, SP-100 reactor design, safety, and testing, materials compatibility issues for fabric composite radiators, application of the enabler to nuclear electric propulsion, orbit-transfer with TOPAZ-type power sources, the thermoelectric properties of alloys, ruthenium silicide as a promising thermoelectric material, and innovative space-saving device for high-temperature piping systems. The second volume of this conference discusses engine concepts for nuclear electric propulsion, nuclear technologies for human exploration of the solar system, dynamic energy conversion, direct nuclear propulsion, thermionic conversion technology, reactor and power system control, thermal management, thermionic research, effects of radiation on electronics, heat-pipe technology, radioisotope power systems, and nuclear fuels for power reactors. The third volume discusses space power electronics, space nuclear fuels for propulsion reactors, power systems concepts, space power electronics systems, the use of artificial intelligence in space, flight qualifications and testing, microgravity two-phase flow, reactor manufacturing and processing, and space and environmental effects.

10 CFR 2.809 - Participation by the Advisory Committee on Reactor Safeguards.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Participation by the Advisory Committee on Reactor Safeguards. 2.809 Section 2.809 Energy NUCLEAR REGULATORY COMMISSION RULES OF PRACTICE FOR DOMESTIC LICENSING PROCEEDINGS AND ISSUANCE OF ORDERS Rulemaking § 2.809 Participation by the Advisory Committee on Reactor...

10 CFR 2.340 - Initial decision in certain contested proceedings; immediate effectiveness of initial decisions...

Code of Federal Regulations, 2010 CFR

2010-01-01

... those matters, the Commission, the Director, Office of Nuclear Reactor Regulation or Director, Office of... Nuclear Reactor Regulation, as appropriate, after making the requisite findings, will issue, deny or... acceptance criteria in nuclear power reactor combined licenses. In any initial decision under § 52.103(g) of...

10 CFR 2.337 - Evidence at a hearing.

Code of Federal Regulations, 2014 CFR

2014-01-01

... Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of... the proceeding under subpart A of part 51 of this chapter by the Director, Office of Nuclear Reactor... 10 Energy 1 2014-01-01 2014-01-01 false Evidence at a hearing. 2.337 Section 2.337 Energy NUCLEAR...

10 CFR 2.340 - Initial decision in certain contested proceedings; immediate effectiveness of initial decisions...

Code of Federal Regulations, 2011 CFR

2011-01-01

... those matters, the Commission, the Director, Office of Nuclear Reactor Regulation or Director, Office of... Nuclear Reactor Regulation, as appropriate, after making the requisite findings, will issue, deny or... acceptance criteria in nuclear power reactor combined licenses. In any initial decision under § 52.103(g) of...

10 CFR 2.340 - Initial decision in certain contested proceedings; immediate effectiveness of initial decisions...

Code of Federal Regulations, 2012 CFR

2012-01-01

... those matters, the Commission, the Director, Office of Nuclear Reactor Regulation or Director, Office of... Nuclear Reactor Regulation, as appropriate, after making the requisite findings, will issue, deny or... acceptance criteria in nuclear power reactor combined licenses. In any initial decision under § 52.103(g) of...

10 CFR 2.337 - Evidence at a hearing.

Code of Federal Regulations, 2013 CFR

2013-01-01

... Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of... the proceeding under subpart A of part 51 of this chapter by the Director, Office of Nuclear Reactor... 10 Energy 1 2013-01-01 2013-01-01 false Evidence at a hearing. 2.337 Section 2.337 Energy NUCLEAR...

Proceedings of the 2013 International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering - M and C 2013

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

2013-07-01

The Mathematics and Computation Division of the American Nuclear (ANS) and the Idaho Section of the ANS hosted the 2013 International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M and C 2013). This proceedings contains over 250 full papers with topics ranging from reactor physics; radiation transport; materials science; nuclear fuels; core performance and optimization; reactor systems and safety; fluid dynamics; medical applications; analytical and numerical methods; algorithms for advanced architectures; and validation verification, and uncertainty quantification.

RADIATION DAMAGE IN REACTOR MATERIALS. Proceedings of the Symposium on Radiation Damage in Solids and Reactor Materials Held in Venice, 7-11 May 1962

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1964-10-31

Thirty papers and 3 reviews of papers and panel discussions presented at the Symposium on Radiation Damage in Solids and Reactor Materials are given. Eighteen papers were previously abstracted for NSA. Separate abstracts were prepared for the remaining 15 papers. (M.C.G.)

Special features of the inverse-beta-decay reaction proceeding on a proton in a reactor-antineutrino flux

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kopeikin, V. I., E-mail: kopeikin46@yandex.ru; Skorokhvatov, M. D., E-mail: skorokhvatov-md@nrcki.ru

2017-03-15

The evolution of the reactor-antineutrino spectrum and the evolution of the spectrum of positrons from the inverse-beta-decay reaction in the course of reactor operation and after reactor shutdown are considered. The present-day status in determining the initial reactor-antineutrino spectrum on the basis of spectra of beta particles from mixtures of products originating from uranium and plutonium fission is described. A local rise of the experimental spectrum of reactor antineutrinos with respect to the expected spectrum is studied.

Space nuclear power systems; Proceedings of the 8th Symposium, Albuquerque, NM, Jan. 6-10, 1991. Pts. 1-3

NASA Astrophysics Data System (ADS)

El-Genk, Mohamed S.; Hoover, Mark D.

1991-07-01

The present conference discusses NASA mission planning for space nuclear power, lunar mission design based on nuclear thermal rockets, inertial-electrostatic confinement fusion for space power, nuclear risk analysis of the Ulysses mission, the role of the interface in refractory metal alloy composites, an advanced thermionic reactor systems design code, and space high power nuclear-pumped lasers. Also discussed are exploration mission enhancements with power-beaming, power requirement estimates for a nuclear-powered manned Mars rover, SP-100 reactor design, safety, and testing, materials compatibility issues for fabric composite radiators, application of the enabler to nuclear electric propulsion, orbit-transfer with TOPAZ-type power sources, the thermoelectric properties of alloys, ruthenium silicide as a promising thermoelectric material, and innovative space-saving device for high-temperature piping systems. The second volume of this conference discusses engine concepts for nuclear electric propulsion, nuclear technologies for human exploration of the solar system, dynamic energy conversion, direct nuclear propulsion, thermionic conversion technology, reactor and power system control, thermal management, thermionic research, effects of radiation on electronics, heat-pipe technology, radioisotope power systems, and nuclear fuels for power reactors. The third volume discusses space power electronics, space nuclear fuels for propulsion reactors, power systems concepts, space power electronics systems, the use of artificial intelligence in space, flight qualifications and testing, microgravity two-phase flow, reactor manufacturing and processing, and space and environmental effects. (For individual items see A93-13752 to A93-13937)

Project of electro-cyclotron resonance ion source test-bench for material investigation.

PubMed

Kulevoy, T V; Chalykh, B B; Kuibeda, R P; Kropachev, G N; Ziiatdinova, A V

2014-02-01

Development of new materials for future energy facilities with higher operating efficiency is a challenging and crucial task. However, full-scale testing of radiation hardness for reactor materials is quite sophisticated and difficult as it requires long session of reactor irradiation; moreover, induced radioactivity considerably complicates further investigation. Ion beam irradiation does not have such a drawback; on the contrary, it has certain advantages. One of them is high speed of defect formation. Therefore, it provides a useful tool for modeling of different radiation damages. Improved understanding of material behavior under high dose irradiation will probably allow to simulate reactor irradiation close to real conditions and to make an adequate estimation of material radiation hardness. Since 2008 in Institute for Theoretical and Experimental Physics, the ion beam irradiation experiments are under development at the heavy ion radio frequency quadrupole linac and very important results are obtained already [T. V. Kulevoy et al., in Proceedings of the International Topical Meeting on Nuclear Research Applications and Utilization of Accelerators, IAEA Vienna, Austria, 2009, http://www.pub.iaea.org/MTCD/publications/PDF/P1433_CD/darasets/papers/ap_p5_07.pdf]. Nevertheless, the new test bench based on electro-cyclotron resonance ion source and high voltage platform is developed. The project of the test bench is presented and discussed.

Project of electro-cyclotron resonance ion source test-bench for material investigation

NASA Astrophysics Data System (ADS)

Kulevoy, T. V.; Chalykh, B. B.; Kuibeda, R. P.; Kropachev, G. N.; Ziiatdinova, A. V.

2014-02-01

Development of new materials for future energy facilities with higher operating efficiency is a challenging and crucial task. However, full-scale testing of radiation hardness for reactor materials is quite sophisticated and difficult as it requires long session of reactor irradiation; moreover, induced radioactivity considerably complicates further investigation. Ion beam irradiation does not have such a drawback; on the contrary, it has certain advantages. One of them is high speed of defect formation. Therefore, it provides a useful tool for modeling of different radiation damages. Improved understanding of material behavior under high dose irradiation will probably allow to simulate reactor irradiation close to real conditions and to make an adequate estimation of material radiation hardness. Since 2008 in Institute for Theoretical and Experimental Physics, the ion beam irradiation experiments are under development at the heavy ion radio frequency quadrupole linac and very important results are obtained already [T. V. Kulevoy et al., in Proceedings of the International Topical Meeting on Nuclear Research Applications and Utilization of Accelerators, IAEA Vienna, Austria, 2009, http://www.pub.iaea.org/MTCD/publications/PDF/P1433_CD/darasets/papers/ap_p5_07.pdf]. Nevertheless, the new test bench based on electro-cyclotron resonance ion source and high voltage platform is developed. The project of the test bench is presented and discussed.

10 CFR 2.102 - Administrative review of application.

Code of Federal Regulations, 2011 CFR

2011-01-01

... completion of its review. (b) The Director, Office of Nuclear Reactor Regulation, Director, Office of New... as requested by the Commission. (c) The Director, Office of Nuclear Reactor Regulation, Director... Energy NUCLEAR REGULATORY COMMISSION RULES OF PRACTICE FOR DOMESTIC LICENSING PROCEEDINGS AND ISSUANCE OF...

10 CFR 2.102 - Administrative review of application.

Code of Federal Regulations, 2012 CFR

2012-01-01

... completion of its review. (b) The Director, Office of Nuclear Reactor Regulation, Director, Office of New... as requested by the Commission. (c) The Director, Office of Nuclear Reactor Regulation, Director... Energy NUCLEAR REGULATORY COMMISSION RULES OF PRACTICE FOR DOMESTIC LICENSING PROCEEDINGS AND ISSUANCE OF...

10 CFR 2.501 - Notice of hearing on application under subpart F of 10 CFR part 52 for a license to manufacture...

Code of Federal Regulations, 2011 CFR

2011-01-01

... part 52 for a license to manufacture nuclear power reactors. 2.501 Section 2.501 Energy NUCLEAR... Procedures Applicable to Proceedings for the Issuance of Licenses To Manufacture Nuclear Power Reactors To Be... power reactors. (a) In the case of an application under subpart F of part 52 of this chapter for a...

10 CFR 2.501 - Notice of hearing on application under subpart F of 10 CFR part 52 for a license to manufacture...

Code of Federal Regulations, 2010 CFR

2010-01-01

... part 52 for a license to manufacture nuclear power reactors. 2.501 Section 2.501 Energy NUCLEAR... Procedures Applicable to Proceedings for the Issuance of Licenses To Manufacture Nuclear Power Reactors To Be... power reactors. (a) In the case of an application under subpart F of part 52 of this chapter for a...

10 CFR 2.102 - Administrative review of application.

Code of Federal Regulations, 2010 CFR

2010-01-01

... Energy NUCLEAR REGULATORY COMMISSION RULES OF PRACTICE FOR DOMESTIC LICENSING PROCEEDINGS AND ISSUANCE OF... proceeding to confer with the NRC staff informally. In the case of docketed application for a limited work... completion of its review. (b) The Director, Office of Nuclear Reactor Regulation, Director, Office of New...

10 CFR 2.502-2.504 - [Reserved

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false [Reserved] 2.502-2.504 Section 2.502-2.504 Energy NUCLEAR REGULATORY COMMISSION RULES OF PRACTICE FOR DOMESTIC LICENSING PROCEEDINGS AND ISSUANCE OF ORDERS Additional Procedures Applicable to Proceedings for the Issuance of Licenses To Manufacture Nuclear Power Reactors To Be...

Proceedings: 2002 Workshop on Pressurized Water Reactor Elevated Feedwater Iron Transport

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

2002-11-01

Some pressurized water reactor (PWR) stations have experienced difficulty with maintaining feedwater (FW) iron concentrations below recommended concentration on a regular basis. A workshop held on September 17-18 in Dana Point, California, addressed the challenge of elevated feedwater iron transport in PWRs.

10 CFR 2.107 - Withdrawal of application.

Code of Federal Regulations, 2011 CFR

2011-01-01

..., Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or Director, Office of Nuclear... 10 Energy 1 2011-01-01 2011-01-01 false Withdrawal of application. 2.107 Section 2.107 Energy NUCLEAR REGULATORY COMMISSION RULES OF PRACTICE FOR DOMESTIC LICENSING PROCEEDINGS AND ISSUANCE OF ORDERS...

10 CFR 2.107 - Withdrawal of application.

Code of Federal Regulations, 2012 CFR

2012-01-01

..., Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or Director, Office of Nuclear... 10 Energy 1 2012-01-01 2012-01-01 false Withdrawal of application. 2.107 Section 2.107 Energy NUCLEAR REGULATORY COMMISSION RULES OF PRACTICE FOR DOMESTIC LICENSING PROCEEDINGS AND ISSUANCE OF ORDERS...

10 CFR 2.500 - Scope of subpart.

Code of Federal Regulations, 2010 CFR

2010-01-01

... Procedures Applicable to Proceedings for the Issuance of Licenses To Manufacture Nuclear Power Reactors To Be... consideration in separate hearings of an application for a license to manufacture nuclear power reactors under... 10 Energy 1 2010-01-01 2010-01-01 false Scope of subpart. 2.500 Section 2.500 Energy NUCLEAR...

Radiation effects and tritium technology for fusion reactors. Volume I. Proceedings of the international conference, Gatlinburg, Tennessee, October 1--3, 1975

DOE Office of Scientific and Technical Information (OSTI.GOV)

Watson, J.S.; Wiffen, F.W.; Bishop, J.L.

1976-03-01

Separate abstracts were prepared for the 29 included papers in Vol. I. The topics covered in this volume include swelling and microstructures in thermonuclear reactor materials. Some papers on modeling and damage analysis are included. (MOW)

10 CFR 2.337 - Evidence at a hearing.

Code of Federal Regulations, 2010 CFR

2010-01-01

... the proffered evidence. If the excluded evidence is in written form, a copy must be marked for... proceeding in compliance with section 182(b) of the Act, any safety evaluation prepared by the NRC staff, and... chapter by the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or...

Hydrogen and water reactor safety: proceedings

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1982-01-01

Separate abstracts were prepared for papers presented in the following areas of interest: 1) hydrogen research programs; 2) hydrogen behavior during light water reactor accidents; 3) combustible gas generation; 4) hydrogen transport and mixing; 5) combustion modeling and experiments; 6) accelerated flames and detonations; 7) combustion mitigation and control; and 8) equipment survivability.

10 CFR 51.105a - Public hearings in proceedings for issuance of manufacturing licenses.

Code of Federal Regulations, 2012 CFR

2012-01-01

... should be issued as proposed by the NRC's Director, Office of New Reactors or Director, Office of Nuclear Reactor Regulation, as appropriate. [73 FR 5724, Jan. 31, 2008] ... manufacturing licenses. 51.105a Section 51.105a Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) ENVIRONMENTAL...

10 CFR 51.105a - Public hearings in proceedings for issuance of manufacturing licenses.

Code of Federal Regulations, 2010 CFR

2010-01-01

... should be issued as proposed by the NRC's Director, Office of New Reactors or Director, Office of Nuclear Reactor Regulation, as appropriate. [73 FR 5724, Jan. 31, 2008] ... manufacturing licenses. 51.105a Section 51.105a Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) ENVIRONMENTAL...

10 CFR 51.105a - Public hearings in proceedings for issuance of manufacturing licenses.

Code of Federal Regulations, 2011 CFR

2011-01-01

... should be issued as proposed by the NRC's Director, Office of New Reactors or Director, Office of Nuclear Reactor Regulation, as appropriate. [73 FR 5724, Jan. 31, 2008] ... manufacturing licenses. 51.105a Section 51.105a Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) ENVIRONMENTAL...

10 CFR 51.105a - Public hearings in proceedings for issuance of manufacturing licenses.

Code of Federal Regulations, 2013 CFR

2013-01-01

... should be issued as proposed by the NRC's Director, Office of New Reactors or Director, Office of Nuclear Reactor Regulation, as appropriate. [73 FR 5724, Jan. 31, 2008] ... manufacturing licenses. 51.105a Section 51.105a Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) ENVIRONMENTAL...

10 CFR 51.105a - Public hearings in proceedings for issuance of manufacturing licenses.

Code of Federal Regulations, 2014 CFR

2014-01-01

... should be issued as proposed by the NRC's Director, Office of New Reactors or Director, Office of Nuclear Reactor Regulation, as appropriate. [73 FR 5724, Jan. 31, 2008] ... manufacturing licenses. 51.105a Section 51.105a Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) ENVIRONMENTAL...

78 FR 2295 - Charlissa C. Smith; Establishment of Atomic Safety and Licensing Board

Federal Register 2010, 2011, 2012, 2013, 2014

2013-01-10

... NUCLEAR REGULATORY COMMISSION [Docket No. 55-23694-SP; ASLBP No. 13-925-01-SP-BD01] Charlissa C... (Board) is being established to preside over the following proceeding: Charlissa C. Smith, (Denial of Senior Reactor Operator License). This proceeding concerns a hearing request from Charlissa C. Smith...

Exploding the myths about the fast breeder reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Burns, S.

1979-01-01

This paper discusses the facts and figures about the effects of conservation policies, the benefits of the Clinch River Breeder Reactor demonstration plant, the feasibility of nuclear weapons manufacture from reactor-grade plutonium, diversion of plutonium from nuclear plants, radioactive waste disposal, and the toxicity of plutonium. The paper concludes that the U.S. is not proceeding with a high confidence strategy for breeder development because of a variety of false assumptions.

10 CFR 2.629 - Finality of partial decision on site suitability issues in a combined license proceeding.

Code of Federal Regulations, 2010 CFR

2010-01-01

... a combined license proceeding. 2.629 Section 2.629 Energy NUCLEAR REGULATORY COMMISSION RULES OF... Work Authorizations Early Partial Decisions on Site Suitability-Combined License Under 10 Cfr Part 52... complete and acceptable for docketing under § 2.101(a)(3), the Director of the Office of New Reactors or...

10 CFR 2.402 - Separate hearings on separate issues; consolidation of proceedings.

Code of Federal Regulations, 2010 CFR

2010-01-01

... appendix N of part 50 of this chapter for construction permits for nuclear power reactors of a type... Licenses To Construct and/or Operate Nuclear Power Plants of Identical Design at Multiple Sites § 2.402... proceedings. 2.402 Section 2.402 Energy NUCLEAR REGULATORY COMMISSION RULES OF PRACTICE FOR DOMESTIC LICENSING...

International Forum on Reactor Ageing Management (IFRAM); Proceedings of the European Engagement Workshop held in Petten, The Netherlands, 25th – 27th May 2010

DOE Office of Scientific and Technical Information (OSTI.GOV)

Martin, Oliver; Bieth, Michel; Bond, Leonard J.

2010-08-14

The report provides the proceedings of the workshop that engaged European representatives to discuss the formation of an International Forum for Reactor Aging Management. There were 29 participants who came from the following countries or international/European Organizations: IAEA, Belgium, Czech Republic, Finland, France, Germany, Hungary, The Netherlands, Romania, Slovakia, Spain, Sweden, Switzerland, Ukraine, USA and EC/JRC. There was a large variety in types of organisations, i.e. utilities, safety authorities, TVOs and research institutes, providing large variety in the presentations. The meeting supported establishing an IFRAM Global Steering Committee and moving forward with the development of this important network.

10 CFR 2.603 - Acceptance and docketing of application for early review of site suitability issues in a...

Code of Federal Regulations, 2010 CFR

2010-01-01

... REGULATORY COMMISSION RULES OF PRACTICE FOR DOMESTIC LICENSING PROCEEDINGS AND ISSUANCE OF ORDERS Additional... Issuance of Limited Work Authorizations Early Partial Decisions on Site Suitability-Construction Permit § 2... processing, the Director of the Office of New Reactors or the Director of the Office of Nuclear Reactor...

10 CFR 2.621 - Acceptance and docketing of application for early review of site suitability issues in a combined...

Code of Federal Regulations, 2010 CFR

2010-01-01

... REGULATORY COMMISSION RULES OF PRACTICE FOR DOMESTIC LICENSING PROCEEDINGS AND ISSUANCE OF ORDERS Additional... Issuance of Limited Work Authorizations Early Partial Decisions on Site Suitability-Combined License Under... Director of the Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation, as...

Characteristics of DO, organic matter, and ammonium profile for practical-scale DHS reactor under various organic load and temperature conditions.

PubMed

Nomoto, Naoki; Ali, Muntjeer; Jayaswal, Komal; Iguchi, Akinori; Hatamoto, Masashi; Okubo, Tsutomu; Takahashi, Masanobu; Kubota, Kengo; Tagawa, Tadashi; Uemura, Shigeki; Yamaguchi, Takashi; Harada, Hideki

2018-04-01

Profile analysis of the down-flow hanging sponge (DHS) reactor was conducted under various temperature and organic load conditions to understand the organic removal and nitrification process for sewage treatment. Under high organic load conditions (3.21-7.89 kg-COD m -3  day -1 ), dissolved oxygen (DO) on the upper layer of the reactor was affected by organic matter concentration and water temperature, and sometimes reaches around zero. Almost half of the COD Cr was removed by the first layer, which could be attributed to the adsorption of organic matter on sponge media. After the first layer, organic removal proceeded along the first-order reaction equation from the second to the fourth layers. The ammoniacal nitrogen removal ratio decreased under high organic matter concentration (above 100 mg L -1 ) and low DO (less than 1 mg L -1 ) condition. Ammoniacal nitrogen removal proceeded via a zero-order reaction equation along the reactor height. In addition, the profile results of DO, COD Cr , and NH 3 -N were different in the horizontal direction. Thus, it is thought the concentration of these items and microbial activities were not in a uniform state even in the same sponge layer of the DHS reactor.

Atomic Safety and Licensing Board Panel Biennial Report, Fiscal Years 1993--1994. Volume 6

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1995-08-01

In Fiscal Year 1993, the Atomic Safety and Licensing Board Panel (``the Panel``) handled 30 proceedings. In Fiscal Year 1994, the Panel handled 36 proceedings. The cases addressed issues in the construction, operation, and maintenance of commercial nuclear power reactors and other activities requiring a license form the Nuclear Regulatory Commission. This report sets out the Panel`s caseload during the year and summarizes, highlight, and analyzes how the wide- ranging issues raised in those proceedings were addressed by the Panel`s judges and licensing boards.

PBF Reactor Building (PER620). Camera facing north toward south facade. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PBF Reactor Building (PER-620). Camera facing north toward south facade. Note west-wing siding on concrete block; high-bay siding of metal. Excavation and forms for signal and cable trenches proceed from building. Photographer: Kirsh. Date August 20, 1968. INEEL negative no. 68-3332 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Proceedings of a Conference on the Role of Education in the Re-Industrialization of the United States (San Francisco, California, March 30, 1980).

ERIC Educational Resources Information Center

Sniegoski, Stephen J., Ed.

These proceedings of a conference on the role of education in the reindustrialization of the United States consist of the conference introductory remarks, the texts of five conference presentations, reactor comments and concluding remarks, an executive summary, and a list of conference participants. The first conference report describes the South…

Japan Report No. 173.

DTIC Science & Technology

1983-06-10

nuclear ship Mutsu . We will in parallel pursue the development of an advanced ma- rine reactor of more compact and ef- ficient design. In the field...velopment include such subjects as the High Temperature Gas Reactor - envisaged for uses other than power generation - and nuclear ship pro- pulsion. We...950’centigrade. In the field of nuclear ship pro- pulsion, we will proceed on our plans for experimental voyages to be undertaken by our first

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 3, Sessions 12-16

DOE Office of Scientific and Technical Information (OSTI.GOV)

Block, R.C.; Feiner, F.

This document, Volume 3, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, ad the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected abstracts have been indexed separately for inclusion in the Energy Science and Technology Database.

Maize mono-digestion efficiency: results from laboratory tests.

PubMed

Ficara, Elena; Malpei, Francesca

2011-01-01

A laboratory experimental campaign was carried out in order to assess the optimal configuration for the anaerobic digestion of a mixture of sweet corn and ensiled maize. Batch hydrolysis tests were conducted at 35 and 55 °C and at four different particle sizes (2, 5, 20 and 50 mm) obtained by manual chopping and sieving. Chemical pre-treatment by 24 h incubation at various acid and alkaline pH was also considered for its potential to increase the maize methane yield. Results suggest that the hydrolytic phase proceeds significantly faster under thermophilic conditions. Significant differences in the solubilization rate were also observed when comparing coarse (20-50 mm) with fine (2-5 mm) particles, while 2 and 5 mm particles were solubilized at similar rates. No advantages from the chemical pre-treatment, in terms of solubilization efficiency and biomethanization potential were observed. According to these preliminary results, a two-stage semi-continuous laboratory plant consisting of a thermophilic hydrolytic reactor followed by a mesophilic methanogenic reactor was operated for 110 days. Steady state loading parameters were: influent concentration (maize mixture diluted in tap water) of 46 g VS/L, hydraulic retention time of 31 d, organic loading rate of 1.5 g VS/L/d. Alkalinity was dosed to the methanogenic reactor to avoid pH drops. Collected data allowed the average biodegradation efficiency to be estimated at around 60-65%.

Heat pulse propagation studies on DIII-D and the Tokamak Fusion Test Reactor

NASA Astrophysics Data System (ADS)

Fredrickson, E. D.; Austin, M. E.; Groebner, R.; Manickam, J.; Rice, B.; Schmidt, G.; Snider, R.

2000-12-01

Sawtooth phenomena have been studied on DIII-D and the Tokamak Fusion Test Reactor (TFTR) [D. Meade and the TFTR Group, in Proceedings of the International Conference on Plasma Physics and Controlled Nuclear Fusion, Washington, DC, 1990 (International Atomic Energy Agency, Vienna, 1991), Vol. 1, pp. 9-24]. In the experiments the sawtooth characteristics were studied with fast electron temperature (ECE) and soft x-ray diagnostics. For the first time, measurements of a strong ballistic electron heat pulse were made in a shaped tokamak (DIII-D) [J. Luxon and DIII-D Group, in Proceedings of the 11th International Conference on Plasma Physics and Controlled Nuclear Fusion Research, Kyoto (International Atomic Energy Agency, Vienna, 1987), Vol. 1, p. 159] and the "ballistic effect" was stronger than was previously reported on TFTR. Evidence is presented in this paper that the ballistic effect is related to the fast growth phase of the sawtooth precursor. Fast, 2 ms interval, measurements on DIII-D were made of the ion temperature evolution following sawteeth and partial sawteeth to document the ion heat pulse characteristics. It is found that the ion heat pulse does not exhibit the very fast, "ballistic" behavior seen for the electrons. Further, for the first time it is shown that the electron heat pulses from partial sawtooth crashes (on DIII-D and TFTR) are seen to propagate at speeds close to those expected from the power balance calculations of the thermal diffusivities whereas heat pulses from fishbones propagate at rates more consistent with sawtooth induced heat pulses. These results suggest that the fast propagation of sawtooth-induced heat pulses is not a feature of nonlinear transport models, but that magnetohydrodynamic events can have a strong effect on electron thermal transport.

Fission Surface Power Technology Development Update

NASA Technical Reports Server (NTRS)

Palac, Donald T.; Mason, Lee S.; Houts, Michael G.; Harlow, Scott

2011-01-01

Power is a critical consideration in planning exploration of the surfaces of the Moon, Mars, and places beyond. Nuclear power is an important option, especially for locations in the solar system where sunlight is limited or environmental conditions are challenging (e.g., extreme cold, dust storms). NASA and the Department of Energy are maintaining the option for fission surface power for the Moon and Mars by developing and demonstrating technology for a fission surface power system. The Fission Surface Power Systems project has focused on subscale component and subsystem demonstrations to address the feasibility of a low-risk, low-cost approach to space nuclear power for surface missions. Laboratory demonstrations of the liquid metal pump, reactor control drum drive, power conversion, heat rejection, and power management and distribution technologies have validated that the fundamental characteristics and performance of these components and subsystems are consistent with a Fission Surface Power preliminary reference concept. In addition, subscale versions of a non-nuclear reactor simulator, using electric resistance heating in place of the reactor fuel, have been built and operated with liquid metal sodium-potassium and helium/xenon gas heat transfer loops, demonstrating the viability of establishing system-level performance and characteristics of fission surface power technologies without requiring a nuclear reactor. While some component and subsystem testing will continue through 2011 and beyond, the results to date provide sufficient confidence to proceed with system level technology readiness demonstration. To demonstrate the system level readiness of fission surface power in an operationally relevant environment (the primary goal of the Fission Surface Power Systems project), a full scale, 1/4 power Technology Demonstration Unit (TDU) is under development. The TDU will consist of a non-nuclear reactor simulator, a sodium-potassium heat transfer loop, a power conversion unit with electrical controls, and a heat rejection system with a multi-panel radiator assembly. Testing is planned at the Glenn Research Center Vacuum Facility 6 starting in 2012, with vacuum and liquid-nitrogen cold walls to provide simulation of operationally relevant environments. A nominal two-year test campaign is planned including a Phase 1 reactor simulator and power conversion test followed by a Phase 2 integrated system test with radiator panel heat rejection. The testing is expected to demonstrate the readiness and availability of fission surface power as a viable power system option for NASA's exploration needs. In addition to surface power, technology development work within this project is also directly applicable to in-space fission power and propulsion systems.

Determination of the Sensitivity of the Antineutrino Probe for Reactor Core Monitoring

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cormon, S.; Fallot, M., E-mail: fallot@subatech.in2p3.fr; Bui, V.-M.

This paper presents a feasibility study of the use of the detection of reactor-antineutrinos (ν{sup ¯}{sub e}) for non proliferation purpose. To proceed, we have started to study different reactor designs with our simulation tools. We use a package called MCNP Utility for Reactor Evolution (MURE), initially developed by CNRS/IN2P3 labs to study Generation IV reactors. The MURE package has been coupled to fission product beta decay nuclear databases for studying reactor antineutrino emission. This method is the only one able to predict the antineutrino emission from future reactor cores, which don't use the thermal fission of {sup 235}U, {supmore » 239}Pu and {sup 241}Pu. It is also the only way to include off-equilibrium effects, due to neutron captures and time evolution of the fission product concentrations during a reactor cycle. We will present here the first predictions of antineutrino energy spectra from innovative reactor designs (Generation IV reactors). We will then discuss a summary of our results of non-proliferation scenarios involving the latter reactor designs, taking into account reactor physics constraints.« less

IECEC '83; Proceedings of the Eighteenth Intersociety Energy Conversion Engineering Conference, Orlando, FL, August 21-26, 1983. Volume 1 - Thermal energy systems

NASA Astrophysics Data System (ADS)

Among the topics discussed are the nuclear fuel cycle, advanced nuclear reactor designs, developments in central status power reactors, space nuclear reactors, magnetohydrodynamic devices, thermionic devices, thermoelectric devices, geothermal systems, solar thermal energy conversion systems, ocean thermal energy conversion (OTEC) developments, and advanced energy conversion concepts. Among the specific questions covered under these topic headings are a design concept for an advanced light water breeder reactor, energy conversion in MW-sized space power systems, directionally solidified cermet electrodes for thermionic energy converters, boron-based high temperature thermoelectric materials, geothermal energy commercialization, solar Stirling cycle power conversion, and OTEC production of methanol. For individual items see A84-30027 to A84-30055

Versatile Affordable Advanced Fuels and Combustion Technologies

DTIC Science & Technology

2010-11-01

2010. 25. M.B. Colket, R.J. Hall, and S.D. Stouffer, ―Modeling Soot Formation in a Well-Stirred Reactor,‖ Paper GT2004-54001, Proceedings of the ASME...Sea and Air, Paper GT2009-59255, June 8-12, 2009. 33. E. Corporan, M.J. DeWitt, C.D. Klingshirn, R. Striebich, and M.-D. Cheng, ―Emissions...Ona, M.J. Wornat, ―Composition Influence on Deposition in Endothermic Fuels,‖ Paper 2006-7973, Proceedings of the 14 th AIAA/AHI Hypersonics

Coal desulfurization in a rotary kiln combustor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cobb, J.T. Jr.

1991-04-22

The focus of our work during the first quarter of 1991 was on combustion tests at the PEDCO rotary kiln reactor at North American Rayon (NARCO) plant in Elizabethton, TN. The tests had essentially tow related objectives: (a) to obtain basic data on the combustion of anthracite culm in a rotary kiln reactor, and (b) upon the test results, determine how best to proceed with our own planned program at the Humphrey Charcoal kiln in Brookville, PA. The rationale for the tests at PEDCO arose from process analysis which posted red flags on the feasibility of burning low-grade, hard-to-burn fuelsmore » like anthracite culms, in the rotary kiln. The PEDCO unit afforded a unique opportunity to obtain some quick answers at low cost. Two different anthracite culm fuels were tested: a so-called Jeddo culm with an average heating value of 7000 Btu/lb, and a relatively poorer culm, and Emerald'' culm, with an average heating value of 5000 Btu/lb. An attempt was also made to burn a blend of the Emerald culm with bituminous coal in 75/25 percent proportions. This report describes the tests, their chronology, and preliminary results. As it turned out, the PEDCO unit is not configured properly for the combustion of anthracite culm. As a result, it proved difficult to achieve a sustained period of steady-state combustion operation, and combustion efficiencies were low even when supplemental fuel was used to aid combustion of the culm. 1 fig., 2 tabs.« less

Institute of Electrical and Electronics Engineers, Nuclear Science Symposium, 18th, and Nuclear Power Systems Symposium, 3rd, San Francisco, Calif., November 3-5, 1971, Proceedings.

NASA Technical Reports Server (NTRS)

1972-01-01

Potential advantages of fusion power reactors are discussed together with the protection of the public from radioactivity produced in nuclear power reactors, and the significance of tritium releases to the environment. Other subjects considered are biomedical instrumentation, radiation damage problems, low level environmental radionuclide analysis systems, nuclear techniques in environmental research, nuclear instrumentation, and space and plasma instrumentation. Individual items are abstracted in this issue.

Thermal evaluation of alternative shipping cask for irradiated experiments

DOE PAGES

Guillen, Donna Post

2015-06-01

Results of a thermal evaluation are provided for a new shipping cask under consideration for transporting irradiated experiments between the test reactor and post-irradiation examination (PIE) facilities. Most of the experiments will be irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL), then later shipped to the Hot Fuel Examination Facility (HFEF) located at the Materials and Fuels Complex for PIE. To date, the General Electric (GE)-2000 cask has been used to transport experiment payloads between these facilities. However, the availability of the GE-2000 cask to support future experiment shipping is uncertain. In addition, the internal cavitymore » of the GE-2000 cask is too short to accommodate shipping the larger payloads. Therefore, an alternate shipping capability is being pursued. The Battelle Energy Alliance, LLC, Research Reactor (BRR) cask has been determined to be the best alternative to the GE-2000 cask. An evaluation of the thermal performance of the BRR cask is necessary before proceeding with fabrication of the newly designed cask hardware and the development of handling, shipping and transport procedures. This paper presents the results of the thermal evaluation of the BRR cask loaded with a representative set of fueled and non-fueled payloads. When analyzed with identical payloads, experiment temperatures were found to be lower with the BRR cask than with the GE-2000 cask. Furthermore, from a thermal standpoint, the BRR cask was found to be a suitable alternate to the GE-2000 cask for shipping irradiated experiment payloads.« less

Application of Reactor Antineutrinos: Neutrinos for Peace

NASA Astrophysics Data System (ADS)

Suekane, F.

2013-02-01

In nuclear reactors, 239Pu are produced along with burn-up of nuclear fuel. 239Pu is subject of safeguard controls since it is an explosive component of nuclear weapon. International Atomic Energy Agency (IAEA) is watching undeclared operation of reactors to prevent illegal production and removal of 239Pu. In operating reactors, a huge numbers of anti electron neutrinos (ν) are produced. Neutrino flux is approximately proportional to the operating power of reactor in short term and long term decrease of the neutrino flux per thermal power is proportional to the amount of 239Pu produced. Thus rector ν's carry direct and real time information useful for the safeguard purposes. Since ν can not be hidden, it could be an ideal medium to monitor the reactor operation. IAEA seeks for novel technologies which enhance their ability and reactor neutrino monitoring is listed as one of such candidates. Currently neutrino physicists are performing R&D of small reactor neutrino detectors to use specifically for the safeguard use in response to the IAEA interest. In this proceedings of the neutrino2012 conference, possibilities of such reactor neutrinos application and current world-wide R&D status are described.

Proceedings of the Flat-Plate Solar Array Workshop on the Science of Silicon Material Preparation

NASA Technical Reports Server (NTRS)

1983-01-01

Several areas of silicon material preparation were addressed including silicon production and purity, thermodynamics, kinetics, mechanisms, particle formation and growth, deposition in fluidized bed reactors, and chemical vapor deposition. Twenty-two papers were presented.

SPERTI/PBF. Contextual aerial view after PBF had begun operating, but ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

SPERT-I/PBF. Contextual aerial view after PBF had begun operating, but prior to expansion of southwest corner of Reactor Building (PER-620). Camera facing northeast. Reactor Building in center of view. Cooling Tower (PER-720) to its left. Warehouse (PER-625) at lower left was built in 1966. SPERT-I Reactor Building (PER-605) and Instrument Cell Building (PER-604) at right of view. Buried cables and piping proceed from PBF toward lower edge of view to Control Building further south and out of view. Photographer: Farmer. Date: March 26, 1976. INEEL negative no. 76-1344 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

An interfaced system for production of methane in a spacecraft

NASA Technical Reports Server (NTRS)

Weiss, A. H.

1973-01-01

The formose reaction, the homogeneously catalyzed condensation of formaldehyde to sugars, proceeds simultaneously with Cannizzaro and crossed Cannizzaro reactions. Reaction studies in a continuous stirred tank reactor have shown that rate instabilities are exhibited. There are temperature instabilities as well as concentration instabilities in calcium hydroxide catalyst, formaldehyde reactant, and hydroxyl ion. It is postulated that Ca(OH)+ is the actual catalytic species for the formose system. A unifying mechanism is developed that postulates that reactions proceed from a common intermediate complexed species, and that the selectivity for each reaction depends on the nature of the catalyst forming the carbohydrate complex. The catalytic mechanism explains the Lobry de Bruyn-van Eckenstein aldose ketose rearrangements and mutarotations of sugars that also proceed in the system.

Runaway Geneeration In Disruptions Of Plasmas In TFTR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fredrickson, E. D.; Bell, M. G.; Taylor, G.

2014-03-31

Many disruptions in the Tokamak Fusion Test Reactor (TFTR) [D. Meade and the TFTR Group, in Proceedings of the International Conference on Plasma Physics and Controlled Nuclear Fusion, Washington, DC, 1990 (International Atomic Energy Agency, Vienna, 1991), Vol. 1, pp. 9-24] produced populations of runaway electrons which carried a significant fraction of the original plasma current. In this paper, we describe experiments where, following a disruption of a low-beta, reversed shear plasma, currents of up to 1 MA carried mainly by runaway electrons were controlled and then ramped down to near zero using the ohmic transformer. In the longer lastingmore » runaway plasmas, Parail-Pogutse instabilities were observed.« less

Decontamination and decommissioning of the Mayaguez (Puerto Rico) facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jackson, P.K.; Freemerman, R.L.

1989-11-01

On February 6, 1987 the US Department of Energy (DOE) awarded the final phase of the decontamination and decommissioning of the nuclear and reactor facilities at the Center for Energy and Environmental Research (CEER), in Mayaguez, Puerto Rico. Bechtel National, Inc., was made the decontamination and decommissioning (D and D) contractor. The goal of the project was to enable DOE to proceed with release of the CEER facility for use by the University of Puerto Rico, who was the operator. This presentation describes that project and lesson learned during its progress. The CEER facility was established in 1957 as themore » Puerto Rico Nuclear Center, a part of the Atoms for Peace Program. It was a nuclear training and research institution with emphasis on the needs of Latin America. It originally consisted of a 1-megawatt Materials Testing Reactor (MTR), support facilities and research laboratories. After eleven years of operation the MTR was shutdown and defueled. A 2-megawatt TRIGA reactor was installed in 1972 and operated until 1976, when it woo was shutdown. Other radioactive facilities at the center included a 10-watt homogeneous L-77 training reactor, a natural uranium graphite-moderated subcritical assembly, a 200KV particle accelerator, and a 15,000 Ci Co-60 irradiation facility. Support facilities included radiochemistry laboratories, counting rooms and two hot cells. As the emphasis shifted to non-nuclear energy technology a name change resulted in the CEER designation, and plans were started for the decontamination and decommissioning effort.« less

10 CFR 2.105 - Notice of proposed action.

Code of Federal Regulations, 2011 CFR

2011-01-01

... NUCLEAR REGULATORY COMMISSION RULES OF PRACTICE FOR DOMESTIC LICENSING PROCEEDINGS AND ISSUANCE OF ORDERS Procedure for Issuance, Amendment, Transfer, or Renewal of a License, and Standard Design Approval Hearing... within the time prescribed in the notice, the Director, Office of Nuclear Reactor Regulation, Director...

10 CFR 2.105 - Notice of proposed action.

Code of Federal Regulations, 2010 CFR

2010-01-01

... NUCLEAR REGULATORY COMMISSION RULES OF PRACTICE FOR DOMESTIC LICENSING PROCEEDINGS AND ISSUANCE OF ORDERS Procedure for Issuance, Amendment, Transfer, or Renewal of a License, and Standard Design Approval Hearing... within the time prescribed in the notice, the Director, Office of Nuclear Reactor Regulation, Director...

PBF Reactor Building (PER620) under construction. Aerial view with camera ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PBF Reactor Building (PER-620) under construction. Aerial view with camera facing northeast. Steel framework is exposed for west wing and high bay. Concrete block siding on east wing. Railroad crane set up on west side. Note trenches proceeding from front of building. Left trench is for secondary coolant and will lead to Cooling Tower. Shorter trench will contain cables leading to control area. Photographer: Larry Page. Date: March 22, 1967. INEEL negative no. 67-5025 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Quench protection diode irradiation tests by the Texas Accelerator Center

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carcagno, R.; Weichold, M.; Welch, G.

1986-07-14

To date considerable progress has been made in the first stage of the TAC program to assess the use of Quench protection diodes (QPD) within the cold region of the SSC superconducting magnets. Our principle goal in this period was to experimentally obtain information that will assist us in designing apparatus for future tests. More specifically, we wished to address the following areas of uncertainty: operational difficulties associated with installing and operating an experiment at a nuclear reactor; diode behavior when subjected to neutron radiation while held close to the temperature of liquid nitrogen; and the extent to which diodesmore » can be annealed by heating them up to 593/degree/K. To gain information in these areas, we performed an irradiation of sixteen Westinghouse diodes at the Texas A and M Nuclear Science Center reactor. In the interest of avoiding the expense and long lead times associated with obtaining new equipment, we used apparatus that was on hand or which could be quickly obtained or constructed. Further tests of the irradiated diodes and analysis of the data already acquired will be continuing for some time, but the results already available will allow us to proceed with designing apparatus for accurate and reliable irradiation testing of diodes. The following is a summary of what we have learned in the three areas mentioned above, some cautions about interpreting the data, and the implications of this new information for our future activities. 14 figs., 1 tab.« less

10 CFR 2.1101 - Purpose.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Purpose. 2.1101 Section 2.1101 Energy NUCLEAR REGULATORY COMMISSION RULES OF PRACTICE FOR DOMESTIC LICENSING PROCEEDINGS AND ISSUANCE OF ORDERS Hybrid Hearing Procedures for Expansion of Spent Nuclear Fuel Storage Capacity at Civilian Nuclear Power Reactors § 2.1101...

10 CFR 2.1113 - Oral argument.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Oral argument. 2.1113 Section 2.1113 Energy NUCLEAR REGULATORY COMMISSION RULES OF PRACTICE FOR DOMESTIC LICENSING PROCEEDINGS AND ISSUANCE OF ORDERS Hybrid Hearing Procedures for Expansion of Spent Nuclear Fuel Storage Capacity at Civilian Nuclear Power Reactors...

10 CFR 2.1113 - Oral argument.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Oral argument. 2.1113 Section 2.1113 Energy NUCLEAR REGULATORY COMMISSION RULES OF PRACTICE FOR DOMESTIC LICENSING PROCEEDINGS AND ISSUANCE OF ORDERS Hybrid Hearing Procedures for Expansion of Spent Nuclear Fuel Storage Capacity at Civilian Nuclear Power Reactors...

10 CFR 2.310 - Selection of hearing procedures.

Code of Federal Regulations, 2010 CFR

2010-01-01

..., licensee-initiated amendment, or termination of licenses or permits for nuclear power reactors, where the... of a civilian nuclear power plant must be conducted under the procedures of subpart L of this part... Energy NUCLEAR REGULATORY COMMISSION RULES OF PRACTICE FOR DOMESTIC LICENSING PROCEEDINGS AND ISSUANCE OF...

Plutonium Recycle: The Fateful Step

ERIC Educational Resources Information Center

Speth, J. Gustave; And Others

1974-01-01

Calls attention to the fact that if the Atomic Energy Commission proceeds with its plans to authorize the nuclear power industry to use plutonium as a fuel in commercial nuclear reactors around the country, this will result in a dramatic escalation in the risks posed by nuclear power. (PEB)

10 CFR 2.1117 - Burden of proof.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Burden of proof. 2.1117 Section 2.1117 Energy NUCLEAR REGULATORY COMMISSION RULES OF PRACTICE FOR DOMESTIC LICENSING PROCEEDINGS AND ISSUANCE OF ORDERS Hybrid Hearing Procedures for Expansion of Spent Nuclear Fuel Storage Capacity at Civilian Nuclear Power Reactors...

10 CFR 2.1117 - Burden of proof.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Burden of proof. 2.1117 Section 2.1117 Energy NUCLEAR REGULATORY COMMISSION RULES OF PRACTICE FOR DOMESTIC LICENSING PROCEEDINGS AND ISSUANCE OF ORDERS Hybrid Hearing Procedures for Expansion of Spent Nuclear Fuel Storage Capacity at Civilian Nuclear Power Reactors...

Microstructural processes in irradiated materials

NASA Astrophysics Data System (ADS)

Byun, Thak Sang; Morgan, Dane; Jiao, Zhijie; Almer, Jonathan; Brown, Donald

2016-04-01

These proceedings contain the papers presented at two symposia, the Microstructural Processes in Irradiated Materials (MPIM) and Characterization of Nuclear Reactor Materials and Components with Neutron and Synchrotron Radiation, held in the TMS 2015, 144th Annual Meeting & Exhibition at Walt Disney World, Orlando, Florida, USA on March 15-19, 2015.

10 CFR 2.400 - Scope of subpart.

Code of Federal Regulations, 2010 CFR

2010-01-01

... to construct and/or operate nuclear power reactors of identical design to be located at multiple... Procedures Applicable to Proceedings for the Issuance of Licenses To Construct and/or Operate Nuclear Power... 10 Energy 1 2010-01-01 2010-01-01 false Scope of subpart. 2.400 Section 2.400 Energy NUCLEAR...

Proceedings of the Conference on High-temperature Electronics

NASA Technical Reports Server (NTRS)

1981-01-01

The development of electronic devices for use in high temperature environments is addressed. The instrumentational needs of planetary exploration, fossil and nuclear power reactors, turbine engine monitoring, and well logging are defined. Emphasis is place on the fabrication and performance of materials and semiconductor devices, circuits and systems and packaging.

Proceedings of GLOBAL 2013: International Nuclear Fuel Cycle Conference - Nuclear Energy at a Crossroads

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

2013-07-01

The Global conference is a forum for the discussion of the scientific, technical, social and regulatory aspects of the nuclear fuel cycle. Relevant topics include global utilization of nuclear energy, current fuel cycle technologies, advanced reactors, advanced fuel cycles, nuclear nonproliferation and public acceptance.

Proceedings of the Joint IAEA/CSNI Specialists` Meeting on Fracture Mechanics Verification by Large-Scale Testing held at Pollard Auditorium, Oak Ridge, Tennessee

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pugh, C.E.; Bass, B.R.; Keeney, J.A.

This report contains 40 papers that were presented at the Joint IAEA/CSNI Specialists` Meeting Fracture Mechanics Verification by Large-Scale Testing held at the Pollard Auditorium, Oak Ridge, Tennessee, during the week of October 26--29, 1992. The papers are printed in the order of their presentation in each session and describe recent large-scale fracture (brittle and/or ductile) experiments, analyses of these experiments, and comparisons between predictions and experimental results. The goal of the meeting was to allow international experts to examine the fracture behavior of various materials and structures under conditions relevant to nuclear reactor components and operating environments. The emphasismore » was on the ability of various fracture models and analysis methods to predict the wide range of experimental data now available. The individual papers have been cataloged separately.« less

Neutron fluxes in test reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Youinou, Gilles Jean-Michel

Communicate the fact that high-power water-cooled test reactors such as the Advanced Test Reactor (ATR), the High Flux Isotope Reactor (HFIR) or the Jules Horowitz Reactor (JHR) cannot provide fast flux levels as high as sodium-cooled fast test reactors. The memo first presents some basics physics considerations about neutron fluxes in test reactors and then uses ATR, HFIR and JHR as an illustration of the performance of modern high-power water-cooled test reactors.

Cross-flow electrochemical reactor cells, cross-flow reactors, and use of cross-flow reactors for oxidation reactions

DOEpatents

Balachandran, Uthamalingam; Poeppel, Roger B.; Kleefisch, Mark S.; Kobylinski, Thaddeus P.; Udovich, Carl A.

1994-01-01

This invention discloses cross-flow electrochemical reactor cells containing oxygen permeable materials which have both electron conductivity and oxygen ion conductivity, cross-flow reactors, and electrochemical processes using cross-flow reactor cells having oxygen permeable monolithic cores to control and facilitate transport of oxygen from an oxygen-containing gas stream to oxidation reactions of organic compounds in another gas stream. These cross-flow electrochemical reactors comprise a hollow ceramic blade positioned across a gas stream flow or a stack of crossed hollow ceramic blades containing a channel or channels for flow of gas streams. Each channel has at least one channel wall disposed between a channel and a portion of an outer surface of the ceramic blade, or a common wall with adjacent blades in a stack comprising a gas-impervious mixed metal oxide material of a perovskite structure having electron conductivity and oxygen ion conductivity. The invention includes reactors comprising first and second zones seprated by gas-impervious mixed metal oxide material material having electron conductivity and oxygen ion conductivity. Prefered gas-impervious materials comprise at least one mixed metal oxide having a perovskite structure or perovskite-like structure. The invention includes, also, oxidation processes controlled by using these electrochemical reactors, and these reactions do not require an external source of electrical potential or any external electric circuit for oxidation to proceed.

Nuclear Science Symposium, 27th, and Symposium on Nuclear Power Systems, 12th, Orlando, Fla., November 5-7, 1980, Proceedings

NASA Technical Reports Server (NTRS)

Martini, M.

1981-01-01

Advances in instrumentation for use in nuclear-science studies are described. Consideration is given to medical instrumentation, computerized fluoroscopy, environmental instrumentation, data acquisition techniques, semiconductor detectors, microchannel plates and photomultiplier tubes, reactor instrumentation, neutron detectors and proportional counters, and space instrumentation.

A mechanism for beam-driven excitation of ion cyclotron harmonic waves in the Tokamak Fusion Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dendy, R.O.; McClements, K.G.; Lashmore-Davies, C.N.

1994-10-01

A mechanism is proposed for the excitation of waves at harmonics of the injected ion cyclotron frequencies in neutral beam-heated discharges in the Tokamak Fusion Test Reactor (TFTR) [[ital Proceedings] [ital of] [ital the] 17[ital th] [ital European] [ital Conference] [ital on] [ital Controlled] [ital Fusion] [ital and] [ital Plasma] [ital Heating] (European Physical Society, Petit-Lancy, Switzerland, 1990), p. 1540]. Such waves are observed to originate from the outer midplane edge of the plasma. It is shown that ion cyclotron harmonic waves can be destabilized by a low concentration of sub-Alfvenic deuterium or tritium beam ions, provided these ions havemore » a narrow distribution of speeds parallel to the magnetic field. Such a distribution is likely to occur in the edge plasma, close to the point of beam injection. The predicted instability gives rise to wave emission at propagation angles lying almost perpendicular to the field. In contrast to the magnetoacoustic cyclotron instability proposed as an excitation mechanism for fusion-product-driven ion cyclotron emission in the Joint European Torus (JET) [Phys. Plasmas [bold 1], 1918 (1994)], the instability proposed here does not involve resonant fast Alfven and ion Bernstein waves, and can be driven by sub-Alfvenic energetic ions. It is concluded that the observed emission from TFTR can be driven by beam ions.« less

Development of a Reactor Model for Chemical Conversion of Lunar Regolith

NASA Technical Reports Server (NTRS)

Hegde, U.; Balasubramaniam, R.; Gokoglu, S.

2009-01-01

Lunar regolith will be used for a variety of purposes such as oxygen and propellant production and manufacture of various materials. The design and development of chemical conversion reactors for processing lunar regolith will require an understanding of the coupling among the chemical, mass and energy transport processes occurring at the length and time scales of the overall reactor with those occurring at the corresponding scales of the regolith particles. To this end, a coupled transport model is developed using, as an example, the reduction of ilmenite-containing regolith by a continuous flow of hydrogen in a flow-through reactor. The ilmenite conversion occurs on the surface and within the regolith particles. As the ilmenite reduction proceeds, the hydrogen in the reactor is consumed, and this, in turn, affects the conversion rate of the ilmenite in the particles. Several important quantities are identified as a result of the analysis. Reactor scale parameters include the void fraction (i.e., the fraction of the reactor volume not occupied by the regolith particles) and the residence time of hydrogen in the reactor. Particle scale quantities include the time for hydrogen to diffuse into the pores of the regolith particles and the chemical reaction time. The paper investigates the relationships between these quantities and their impact on the regolith conversion. Application of the model to various chemical reactor types, such as fluidized-bed, packed-bed, and rotary-bed configurations, are discussed.

Development of a Reactor Model for Chemical Conversion of Lunar Regolith

NASA Technical Reports Server (NTRS)

Hedge, uday; Balasubramaniam, R.; Gokoglu, S.

2007-01-01

Lunar regolith will be used for a variety of purposes such as oxygen and propellant production and manufacture of various materials. The design and development of chemical conversion reactors for processing lunar regolith will require an understanding of the coupling among the chemical, mass and energy transport processes occurring at the length and time scales of the overall reactor with those occurring at the corresponding scales of the regolith particles. To this end, a coupled transport model is developed using, as an example, the reduction of ilmenite-containing regolith by a continuous flow of hydrogen in a flow-through reactor. The ilmenite conversion occurs on the surface and within the regolith particles. As the ilmenite reduction proceeds, the hydrogen in the reactor is consumed, and this, in turn, affects the conversion rate of the ilmenite in the particles. Several important quantities are identified as a result of the analysis. Reactor scale parameters include the void fraction (i.e., the fraction of the reactor volume not occupied by the regolith particles) and the residence time of hydrogen in the reactor. Particle scale quantities include the time for hydrogen to diffuse into the pores of the regolith particles and the chemical reaction time. The paper investigates the relationships between these quantities and their impact on the regolith conversion. Application of the model to various chemical reactor types, such as fluidized-bed, packed-bed, and rotary-bed configurations, are discussed.

Quantification of syntrophic acetate-oxidizing microbial communities in biogas processes

PubMed Central

Westerholm, Maria; Dolfing, Jan; Sherry, Angela; Gray, Neil D; Head, Ian M; Schnürer, Anna

2011-01-01

Changes in communities of syntrophic acetate-oxidizing bacteria (SAOB) and methanogens caused by elevated ammonia levels were quantified in laboratory-scale methanogenic biogas reactors operating at moderate temperature (37°C) using quantitative polymerase chain reaction (qPCR). The experimental reactor was subjected to gradually increasing ammonia levels (0.8–6.9 g NH4+-N l−1), whereas the level of ammonia in the control reactor was kept low (0.65–0.90 g NH4+-N l−1) during the entire period of operation (660 days). Acetate oxidation in the experimental reactor, indicated by increased production of 14CO2 from acetate labelled in the methyl carbon, occurred when ammonia levels reached 5.5 and 6.9 g NH4+-N l−1. Syntrophic acetate oxidizers targeted by newly designed qPCR primers were Thermacetogenium phaeum, Clostridium ultunense, Syntrophaceticus schinkii and Tepidanaerobacter acetatoxydans. The results showed a significant increase in abundance of all these bacteria except T. phaeum in the ammonia-stressed reactor, coincident with the shift to syntrophic acetate oxidation. As the abundance of the bacteria increased, a simultaneous decrease was observed in the abundance of aceticlastic methanogens from the families Methanosaetaceae and Methanosarcinaceae. qPCR analyses of sludge from two additional high ammonia processes, in which methane production from acetate proceeded through syntrophic acetate oxidation (reactor SB) or through aceticlastic degradation (reactor DVX), demonstrated that SAOB were significantly more abundant in the SB reactor than in the DVX reactor. PMID:23761313

FSA future directions: FSA technology activities in FY86

NASA Technical Reports Server (NTRS)

Leipold, M. H.

1985-01-01

The silicon material, advanced silicon sheet, device research, and process research activities are explained. There will be no new initiatives. Many activities are targeted for completion and the emphasis will then be on technology transfer. Industrial development of the fluidized-bed reactor (FBR) deposition technology is proceeding. Technology transfer and industry funding of sheet development are continuing.

Thermophilic treatment of acidified and partially acidified wastewater using an anaerobic submerged MBR: Factors affecting long-term operational flux.

PubMed

Jeison, D; van Lier, J B

2007-09-01

The long-term operation of two thermophilic anaerobic submerged membrane bioreactors (AnSMBRs) was studied using acidified and partially acidified synthetic wastewaters. In both reactors, cake formation was identified as the key factor governing critical flux. Even though cake formation was observed to be mostly reversible, particle deposition proceeds fast once the critical flux is exceeded. Very little irreversible fouling was observed during long-term operation, irrespective of the substrate. Critical flux values at the end of the reactors operation were 7 and 3L/m(2)h for the AnSMBRs fed with acidified and partially acidified wastewaters, respectively, at a gas superficial velocity of 70m/h. Small particle size was identified as the responsible parameter for the low observed critical flux values. The degree of wastewater acidification significantly affected the physical properties of the sludge, determining the attainable flux. Based on the fluxes observed in this research, the membrane costs would be in the range of 0.5euro/m(3) of treated wastewater. Gas sparging was ineffective in increasing the critical flux values. However, preliminary tests showed that cross-flow operation may be a feasible alternative to reduce particle deposition.

Project Execution Plan for the Remote Handled Low-Level Waste Disposal Project

DOE Office of Scientific and Technical Information (OSTI.GOV)

Danny Anderson

2014-07-01

As part of ongoing cleanup activities at the Idaho National Laboratory (INL), closure of the Radioactive Waste Management Complex (RWMC) is proceeding under the Comprehensive Environmental Response, Compensation, and Liability Act (42 USC 9601 et seq. 1980). INL-generated radioactive waste has been disposed of at RWMC since 1952. The Subsurface Disposal Area (SDA) at RWMC accepted the bulk of INL’s contact and remote-handled low-level waste (LLW) for disposal. Disposal of contact-handled LLW and remote-handled LLW ion-exchange resins from the Advanced Test Reactor in the open pit of the SDA ceased September 30, 2008. Disposal of remote-handled LLW in concrete disposalmore » vaults at RWMC will continue until the facility is full or until it must be closed in preparation for final remediation of the SDA (approximately at the end of fiscal year FY 2017). The continuing nuclear mission of INL, associated ongoing and planned operations, and Naval spent fuel activities at the Naval Reactors Facility (NRF) require continued capability to appropriately dispose of contact and remote handled LLW. A programmatic analysis of disposal alternatives for contact and remote-handled LLW generated at INL was conducted by the INL contractor in Fiscal Year 2006; subsequent evaluations were completed in Fiscal Year 2007. The result of these analyses was a recommendation to the Department of Energy (DOE) that all contact-handled LLW generated after September 30, 2008, be disposed offsite, and that DOE proceed with a capital project to establish replacement remote-handled LLW disposal capability. An analysis of the alternatives for providing replacement remote-handled LLW disposal capability has been performed to support Critical Decision-1. The highest ranked alternative to provide this required capability has been determined to be the development of a new onsite remote-handled LLW disposal facility to replace the existing remote-handled LLW disposal vaults at the SDA. Several offsite DOE and commercial disposal options exist for contact-handled LLW; however, offsite disposal options are either not currently available (i.e., commercial disposal facilities), practical, or cost-effective for all remote-handled LLW streams generated at INL. Offsite disposal of all INL and tenant-generated remote-handled waste is further complicated by issues associated with transporting highly radioactive waste in commerce; and infrastructure and processing changes at the generating facilities, specifically NRF, that would be required to support offsite disposal. The INL Remote-Handled LLW Disposal Project will develop a new remote handled LLW disposal facility to meet mission-critical, remote-handled LLW disposal needs. A formal DOE decision to proceed with the project has been made in accordance with the requirements of National Environmental Policy Act (42 USC§ 4321 et seq.). Remote-handled LLW is generated from nuclear programs conducted at INL, including spent nuclear fuel handling and operations at NRF and operations at the Advanced Test Reactor. Remote-handled LLW also will be generated by new INL programs and from segregation and treatment (as necessary) of remote handled scrap and waste currently stored in the Radioactive Scrap and Waste Facility at the Materials and Fuels Complex.« less

KINETICS OF TREAT USED AS A TEST REACTOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dickerman, C.E.; Johnson, R.D.; Gasidlo, J.

1962-05-01

An analysis is presented concerning the reactor kinetics of TREAT used as a pulsed, engineering test reactor for fast reactor fuel element studies. A description of the reactor performance is given for a wide range of conditions associated with its use as a test reactor. Supplemental information on meltdown experimentation is included. (J.R.D.)

Overview of International Thermonuclear Experimental Reactor (ITER) engineering design activities*

NASA Astrophysics Data System (ADS)

Shimomura, Y.

1994-05-01

The International Thermonuclear Experimental Reactor (ITER) [International Thermonuclear Experimental Reactor (ITER) (International Atomic Energy Agency, Vienna, 1988), ITER Documentation Series, No. 1] project is a multiphased project, presently proceeding under the auspices of the International Atomic Energy Agency according to the terms of a four-party agreement among the European Atomic Energy Community (EC), the Government of Japan (JA), the Government of the Russian Federation (RF), and the Government of the United States (US), ``the Parties.'' The ITER project is based on the tokamak, a Russian invention, and has since been brought to a high level of development in all major fusion programs in the world. The objective of ITER is to demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes. The ITER design is being developed, with support from the Parties' four Home Teams and is in progress by the Joint Central Team. An overview of ITER Design activities is presented.

Computers in Composition Instruction. The Proceedings of a Research/Practice Conference Held at SWRL Educational Research and Development (Los Alamitos, California, April 22-23, 1982).

ERIC Educational Resources Information Center

Lawlor, Joseph, Ed.

Suggestions for integrating computer technology and composition instruction are presented in four conference papers, summaries of four conference courseware demonstrations, a paper describing computer-based evaluation of textual responses, and a reactor's address. In an overview of the current state of computer-based composition instruction,…

1986 Nuclear Science Symposium, 33rd, and 1986 Symposium on Nuclear Power Systems, 18th, Washington, DC, Oct. 29-31, 1986, Proceedings

NASA Technical Reports Server (NTRS)

Stubblefield, F. W. (Editor)

1987-01-01

Papers are presented on space, low-energy physics, and general nuclear science instrumentations. Topics discussed include data acquisition systems and circuits, nuclear medicine imaging and tomography, and nuclear radiation detectors. Consideration is given to high-energy physics instrumentation, reactor systems and safeguards, health physics instrumentation, and nuclear power systems.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Aaron, Adam M.; Cunningham, Richard Burns; Fugate, David L.

Effective high-temperature thermal energy exchange and delivery at temperatures over 600°C has the potential of significant impact by reducing both the capital and operating cost of energy conversion and transport systems. It is one of the key technologies necessary for efficient hydrogen production and could potentially enhance efficiencies of high-temperature solar systems. Today, there are no standard commercially available high-performance heat transfer fluids above 600°C. High pressures associated with water and gaseous coolants (such as helium) at elevated temperatures impose limiting design conditions for the materials in most energy systems. Liquid salts offer high-temperature capabilities at low vapor pressures, goodmore » heat transport properties, and reasonable costs and are therefore leading candidate fluids for next-generation energy production. Liquid-fluoride-salt-cooled, graphite-moderated reactors, referred to as Fluoride Salt Reactors (FHRs), are specifically designed to exploit the excellent heat transfer properties of liquid fluoride salts while maximizing their thermal efficiency and minimizing cost. The FHR s outstanding heat transfer properties, combined with its fully passive safety, make this reactor the most technologically desirable nuclear power reactor class for next-generation energy production. Multiple FHR designs are presently being considered. These range from the Pebble Bed Advanced High Temperature Reactor (PB-AHTR) [1] design originally developed by UC-Berkeley to the Small Advanced High-Temperature Reactor (SmAHTR) and the large scale FHR both being developed at ORNL [2]. The value of high-temperature, molten-salt-cooled reactors is also recognized internationally, and Czechoslovakia, France, India, and China all have salt-cooled reactor development under way. The liquid salt experiment presently being developed uses the PB-AHTR as its focus. One core design of the PB-AHTR features multiple 20 cm diameter, 3.2 m long fuel channels with 3 cm diameter graphite-based fuel pebbles slowly circulating up through the core. Molten salt coolant (FLiBe) at 700°C flows concurrently (at significantly higher velocity) with the pebbles and is used to remove heat generated in the reactor core (approximately 1280 W/pebble), and supply it to a power conversion system. Refueling equipment continuously sorts spent fuel pebbles and replaces spent or damaged pebbles with fresh fuel. By combining greater or fewer numbers of pebble channel assemblies, multiple reactor designs with varying power levels can be offered. The PB-AHTR design is discussed in detail in Reference [1] and is shown schematically in Fig. 1. Fig. 1. PB-AHTR concept (drawing taken from Peterson et al., Design and Development of the Modular PB-AHTR Proceedings of ICApp 08). Pebble behavior within the core is a key issue in proving the viability of this concept. This includes understanding the behavior of the pebbles thermally, hydraulically, and mechanically (quantifying pebble wear characteristics, flow channel wear, etc). The experiment being developed is an initial step in characterizing the pebble behavior under realistic PB-AHTR operating conditions. It focuses on thermal and hydraulic behavior of a static pebble bed using a convective salt loop to provide prototypic fluid conditions to the bed, and a unique inductive heating technique to provide prototypic heating in the pebbles. The facility design is sufficiently versatile to allow a variety of other experimentation to be performed in the future. The facility can accommodate testing of scaled reactor components or sub-components such as flow diodes, salt-to-salt heat exchangers, and improved pump designs as well as testing of refueling equipment, high temperature instrumentation, and other reactor core designs.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wheeler, F.; Wessol, D.; Atkinson, C.

During the past few years, murine and large animal research, as well as human studies have provided data to the point where human clinical trials have been initiated at the BMRR using BPA-F for gliomas and at the Massachusetts Institute of Technology Reactor (MITR) using BPA for melanomas of the extremeties. It is expected that glioma trials using BSH will proceed soon at the Petten High Flux Reactor (HFR) in the Netherlands. The first human glioma epithermal boron neutron capture therapy application was performed at the BMRR in the fall of 1994. This was a collaborative effort by BNL, Bethmore » Israel Manhattan hospital, and INEL. The INEL planning system was chosen to perform dose predictions for this application.« less

An experimental test plan for the characterization of molten salt thermochemical properties in heat transport systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pattrick Calderoni

2010-09-01

Molten salts are considered within the Very High Temperature Reactor program as heat transfer media because of their intrinsically favorable thermo-physical properties at temperatures starting from 300 C and extending up to 1200 C. In this context two main applications of molten salt are considered, both involving fluoride-based materials: as primary coolants for a heterogeneous fuel reactor core and as secondary heat transport medium to a helium power cycle for electricity generation or other processing plants, such as hydrogen production. The reference design concept here considered is the Advanced High Temperature Reactor (AHTR), which is a large passively safe reactormore » that uses solid graphite-matrix coated-particle fuel (similar to that used in gas-cooled reactors) and a molten salt primary and secondary coolant with peak temperatures between 700 and 1000 C, depending upon the application. However, the considerations included in this report apply to any high temperature system employing fluoride salts as heat transfer fluid, including intermediate heat exchangers for gas-cooled reactor concepts and homogenous molten salt concepts, and extending also to fast reactors, accelerator-driven systems and fusion energy systems. The purpose of this report is to identify the technical issues related to the thermo-physical and thermo-chemical properties of the molten salts that would require experimental characterization in order to proceed with a credible design of heat transfer systems and their subsequent safety evaluation and licensing. In particular, the report outlines an experimental R&D test plan that would have to be incorporated as part of the design and operation of an engineering scaled facility aimed at validating molten salt heat transfer components, such as Intermediate Heat Exchangers. This report builds on a previous review of thermo-physical properties and thermo-chemical characteristics of candidate molten salt coolants that was generated as part of the same project [1]. However, this work focuses on two materials: the LiF-BeF2 eutectic (67 and 33 mol%, respectively, also known as flibe) as primary coolant and the LiF-NaF-KF eutectic (46.5, 11.5, and 52 mol%, respectively, also known as flinak) as secondary heat transport fluid. At first common issues are identified, involving the preparation and purification of the materials as well as the development of suitable diagnostics. Than issues specific to each material and its application are considered, with focus on the compatibility with structural materials and the extension of the existing properties database.« less

Advanced Test Reactor Tour

ScienceCinema

Miley, Don

2017-12-21

The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored.

Continuous flow synthesis of ZSM-5 zeolite on the order of seconds

PubMed Central

Liu, Zhendong; Okabe, Kotatsu; Anand, Chokkalingam; Yonezawa, Yasuo; Zhu, Jie; Yamada, Hiroki; Endo, Akira; Yanaba, Yutaka; Yoshikawa, Takeshi; Ohara, Koji; Okubo, Tatsuya; Wakihara, Toru

2016-01-01

The hydrothermal synthesis of zeolites carried out in batch reactors takes a time so long (typically, on the order of days) that the crystallization of zeolites has long been believed to be very slow in nature. We herein present a synthetic process for ZSM-5, an industrially important zeolite, on the order of seconds in a continuous flow reactor using pressurized hot water as a heating medium. Direct mixing of a well-tuned precursor (90 °C) with the pressurized water preheated to extremely high temperature (370 °C) in the millimeter-sized continuous flow reactor resulted in immediate heating to high temperatures (240–300 °C); consequently, the crystallization of ZSM-5 in a seed-free system proceeded to completion within tens of or even several seconds. These results indicate that the crystallization of zeolites can complete in a period on the order of seconds. The subtle design combining a continuous flow reactor with pressurized hot water can greatly facilitate the mass production of zeolites in the future. PMID:27911823

Microstructural processes in irradiated materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Byun, Thak Sang; Morgan, Dane; Jiao, Zhijie

2016-04-01

This is an editorial article (preface) for the publication of symposium papers in the Journal of Nuclear materials: These proceedings contain the papers presented at two symposia, the Microstructural Processes in Irradiated Materials (MPIM) and Characterization of Nuclear Reactor Materials and Components with Neutron and Synchrotron Radiation, held in the TMS 2015, 144th Annual Meeting & Exhibition at Walt Disney World, Orlando, Florida, USA on March 15–19, 2015.

10 CFR 2.401 - Notice of hearing on construction permit or combined license applications pursuant to appendix N...

Code of Federal Regulations, 2010 CFR

2010-01-01

... pursuant to appendix N of part 50 of this chapter for construction permits for nuclear power reactors of... Procedures Applicable to Proceedings for the Issuance of Licenses To Construct and/or Operate Nuclear Power... license applications pursuant to appendix N of 10 CFR parts 50 or 52. 2.401 Section 2.401 Energy NUCLEAR...

An exceptionally rapid and selective hydrogenation of 2-cyclohexen-1-one in supercritical carbon dioxide.

PubMed

Chatterjee, Maya; Yokoyama, Toshirou; Kawanami, Hajime; Sato, Masahiro; Suzuki, Toshishige

2009-02-14

Selective hydrogenation of 2-cyclohexen-1-one over Pt-MCM-41 proceeds at a very high rate and produces cyclohexanone with selectivity of 100% in a batch reactor; a marked increase in the reaction rate (TOF) from 2283 min(-1) to 5051 min(-1) is observed on increasing the pressure from 7 MPa to 14 MPa at 40 degrees C.

Next generation fuel irradiation capability in the High Flux Reactor Petten

NASA Astrophysics Data System (ADS)

Fütterer, Michael A.; D'Agata, Elio; Laurie, Mathias; Marmier, Alain; Scaffidi-Argentina, Francesco; Raison, Philippe; Bakker, Klaas; de Groot, Sander; Klaassen, Frodo

2009-07-01

This paper describes selected equipment and expertise on fuel irradiation testing at the High Flux Reactor (HFR) in Petten, The Netherlands. The reactor went critical in 1961 and holds an operating license up to at least 2015. While HFR has initially focused on Light Water Reactor fuel and materials, it also played a decisive role since the 1970s in the German High Temperature Reactor (HTR) development program. A variety of tests related to fast reactor development in Europe were carried out for next generation fuel and materials, in particular for Very High Temperature Reactor (V/HTR) fuel, fuel for closed fuel cycles (U-Pu and Th-U fuel cycle) and transmutation, as well as for other innovative fuel types. The HFR constitutes a significant European infrastructure tool for the development of next generation reactors. Experimental facilities addressed include V/HTR fuel tests, a coated particle irradiation rig, and tests on fast reactor, transmutation and thorium fuel. The rationales for these tests are given, results are provided and further work is outlined.

A facility for testing 10 to 100-kWe space power reactors

NASA Astrophysics Data System (ADS)

Carlson, William F.; Bitten, Ernest J.

1993-01-01

This paper describes an existing facility that could be used in a cost-effective manner to test space power reactors in the 10 to 100-kWe range before launch. The facility has been designed to conduct full power tests of 100-kWe SP-100 reactor systems and already has the structural features that would be required for lower power testing. The paper describes a reasonable scenario starting with the acceptance at the test site of the unfueled reactor assembly and the separately shipped nuclear fuel. After fueling the reactor and installing it in the facility, cold critical tests are performed, and the reactor is then shipped to the launch site. The availability of this facility represents a cost-effective means of performing the required prelaunch test program.

Irradiation Tests Supporting LEU Conversion of Very High Power Research Reactors in the US

DOE Office of Scientific and Technical Information (OSTI.GOV)

Woolstenhulme, N. E.; Cole, J. I.; Glagolenko, I.

The US fuel development team is developing a high density uranium-molybdenum alloy monolithic fuel to enable conversion of five high-power research reactors. Previous irradiation tests have demonstrated promising behavior for this fuel design. A series of future irradiation tests will enable selection of final fuel fabrication process and provide data to qualify the fuel at moderately-high power conditions for use in three of these five reactors. The remaining two reactors, namely the Advanced Test Reactor and High Flux Isotope Reactor, require additional irradiation tests to develop and demonstrate the fuel’s performance with even higher power conditions, complex design features, andmore » other unique conditions. This paper reviews the program’s current irradiation testing plans for these moderately-high irradiation conditions and presents conceptual testing strategies to illustrate how subsequent irradiation tests will build upon this initial data package to enable conversion of these two very-high power research reactors.« less

TEST REACTOR AREA PLOT PLAN CA. 1968. MTR AND ETR ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

TEST REACTOR AREA PLOT PLAN CA. 1968. MTR AND ETR AREAS SOUTH OF PERCH AVENUE. "COLD" SERVICES NORTH OF PERCH. ADVANCED TEST REACTOR IN NEW SECTION WEST OF COLD SERVICES SECTION. NEW PERIMETER FENCE ENCLOSES BETA RAY SPECTROMETER, TRA-669, AN ATR SUPPORT FACILITY, AND ATR STACK. UTM LOCATORS HAVE BEEN DELETED. IDAHO NUCLEAR CORPORATION, FROM A BLAW-KNOX DRAWING, 3/1968. INL INDEX NO. 530-0100-00-400-011646, REV. 0. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Investigation of the reaction of liquid hydrogen with liquid air in a pressure tube

NASA Technical Reports Server (NTRS)

Karb, Erich H.

1987-01-01

A pressure tube should protect the FR-2 reactor from the consequences of a hydrogen-air reaction, which is conceivable in the breakdown of several safety devices of the planned cold neutron source Project FR-2/16. The magnitudes and time pattern of the pressures to be expected were investigated. In the geometry used and the ignition mechanism selected, which is comparable to the strongest ignition process conceivable in the reactor, the reaction proceeds with greater probability than combustion. The combustion is possibly smaller if local limited partial detonations are superimposed. The magnitude of the pressure was determined by the masses of the reaction partners, liquid H2 and liquid air, and determines their ratio to each other.

Summary of NR Program Prometheus Efforts

DOE Office of Scientific and Technical Information (OSTI.GOV)

J Ashcroft; C Eshelman

2006-02-08

The Naval Reactors Program led work on the development of a reactor plant system for the Prometheus space reactor program. The work centered on a 200 kWe electric reactor plant with a 15-20 year mission applicable to nuclear electric propulsion (NEP). After a review of all reactor and energy conversion alternatives, a direct gas Brayton reactor plant was selected for further development. The work performed subsequent to this selection included preliminary nuclear reactor and reactor plant design, development of instrumentation and control techniques, modeling reactor plant operational features, development and testing of core and plant material options, and development ofmore » an overall project plan. Prior to restructuring of the program, substantial progress had been made on defining reference plant operating conditions, defining reactor mechanical, thermal and nuclear performance, understanding the capabilities and uncertainties provided by material alternatives, and planning non-nuclear and nuclear system testing. The mission requirements for the envisioned NEP missions cannot be accommodated with existing reactor technologies. Therefore concurrent design, development and testing would be needed to deliver a functional reactor system. Fuel and material performance beyond the current state of the art is needed. There is very little national infrastructure available for fast reactor nuclear testing and associated materials development and testing. Surface mission requirements may be different enough to warrant different reactor design approaches and development of a generic multi-purpose reactor requires substantial sacrifice in performance capability for each mission.« less

Proceedings of the Natick Science Symposium (4th) Held in Natick, Massachusetts on 9-10 June 1992

DTIC Science & Technology

1992-09-01

HQ - hydroquinone; MG - methyl gallate; CA - caffeic acid ; CHIA - chlorogenic acid ; GA - gallic acid ; FOLY - polymeric antioxidant. Values shown...examples 268 FQRrER & BIACK (qu~tin, hydroquinone, caffeic acid , chlorogenic acid and gallic acid ) , these are relatively ineffective in HSV...unless covalently bonded to the membrane. 279 SENECAL & RAND A membrane lactase reactor, modelled after the mammalian small intestine, has been

Status and progress of the RERTR program in the year 2000.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Travelli, A.

2000-09-28

This paper describes the progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners during the year 2000 and discusses the main activities planned for the year 2001. The past year was characterized by important accomplishments and events for the RERTR program. Four additional shipments containing 503 spent fuel assemblies from foreign research reactors were accepted by the U.S. Altogether, 3,740 spent fuel assemblies from foreign research reactors have been received by the U.S. under the acceptance policy. Postirradiation examinations of three batches of microplates have continued to reveal excellentmore » irradiation behavior of U-MO dispersion fuels in a variety of compositions and irradiating conditions. h-radiation of two new batches of miniplates of greater sizes is in progress in the ATR to investigate me swelling behavior of these fuels under prototypic conditions. These materials hold the promise of achieving the program goal of developing LEU research reactor fuels with uranium densities in the 8-9 g /cm{sup 3} range. Qualification of the U-MO dispersion fuels is proceeding on schedule. Test fuel elements with 6 gU/cm{sup 3} are being fabricated by BWXT and are scheduled to begin undergoing irradiation in the HFR-Petten in the spring of 2001, with a goal of qualifying this fuel by the end of 2003. U-Mo with 8-9 gU/cm{sup 3} is planned to be qualified by the end of 2005. Joint LEU conversion feasibility studies were completed for HFR-Petten and for SAFARI-1. Significant improvements were made in the design of LEU metal-foil annular targets that would allow efficient production of fission {sup 99}Mo. Irradiations in the RAS-GAS reactor showed that these targets can formed from aluminum tubes, and that the yield and purity of their product from the acidic process were at least as good as those from the HEU Cintichem targets. Progress was made on irradiation testing of LEU UO{sub 2} dispersion fuel and on LEU conversion feasibility studies in the Russian RERTR program. Conversion of the BER-11reactor in Berlin, Germany, was completed and conversion of the La Reins reactor in Santiago, Chile, began. These are exciting times for the program. In the fuel development area, the RERTR program is aggressively pursuing qualification of high-density LEU U-Mo dispersion fuels, with the dual goal of enabling fi.uther conversions and of developing a substitute for LEU silicide fuels that can be more easily disposed of after expiration of the FRR SNF Acceptance Program. The {sup 99}Mo effort has reached the point where it appears feasible for all the {sup 99}Mo producers of the world to agree jointly to a common course of action leading to the elimination of HEU use in their processes. As in the past, the success of the RERTR program will depend on the international friendship and cooperation that has always been its trademark.« less

Irradiation Testing Vehicles for Fast Reactors from Open Test Assemblies to Closed Loops

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sienicki, James J.; Grandy, Christopher

A review of irradiation testing vehicle approaches and designs that have been incorporated into past Sodium-Cooled Fast Reactors (SFRs) or envisioned for incorporation has been carried out. The objective is to understand the essential features of the approaches and designs so that they can inform test vehicle designs for a future U.S. Fast Test Reactor. Fast test reactor designs examined include EBR-II, FFTF, JOYO, BOR-60, PHÉNIX, JHR, and MBIR. Previous designers exhibited great ingenuity in overcoming design and operational challenges especially when the original reactor plant’s mission changed to an irradiation testing mission as in the EBRII reactor plant. Themore » various irradiation testing vehicles can be categorized as: Uninstrumented open assemblies that fit into core locations; Instrumented open test assemblies that fit into special core locations; Self-contained closed loops; and External closed loops. A special emphasis is devoted to closed loops as they are regarded as a very desirable feature of a future U.S. Fast Test Reactor. Closed loops are an important technology for irradiation of fuels and materials in separate controlled environments. The impact of closed loops on the design of fast reactors is also discussed in this report.« less

Test Results from a Direct Drive Gas Reactor Simulator Coupled to a Brayton Power Conversion Unit

NASA Technical Reports Server (NTRS)

Hervol, David S.; Briggs, Maxwell H.; Owen, Albert K.; Bragg-Sitton, Shannon M.; Godfroy, Thomas J.

2010-01-01

Component level testing of power conversion units proposed for use in fission surface power systems has typically been done using relatively simple electric heaters for thermal input. These heaters do not adequately represent the geometry or response of proposed reactors. As testing of fission surface power systems transitions from the component level to the system level it becomes necessary to more accurately replicate these reactors using reactor simulators. The Direct Drive Gas-Brayton Power Conversion Unit test activity at the NASA Glenn Research Center integrates a reactor simulator with an existing Brayton test rig. The response of the reactor simulator to a change in Brayton shaft speed is shown as well as the response of the Brayton to an insertion of reactivity, corresponding to a drum reconfiguration. The lessons learned from these tests can be used to improve the design of future reactor simulators which can be used in system level fission surface power tests.

Loss-of-flow-without-scram tests in Experimental Breeder Reactor-II and comparison with pretest predictions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chang, L.K.; Mohr, D.; Planchon, H.P.

This article discusses a series of successful loss-of-flow-without-scram tests conducted in Experimental Breeder Reactor-II (EBR-II), a metal-fueled, sodium-cooled fast reactor. These May 1985 tests demonstrated the capability of the EBR to reduce reactor power passively during a loss of flow and to maintain reactor temperatures within bounds without any reliance on an active safety system. The tests were run from reduced power to ensure that temperatures could be maintained well below the fuel-clad eutectic temperature. Good agreement was found between selected test data and pretest predictions made with the EBR-II system analysis code NATDEMO and the hot channel analysis codemore » HOTCHAN. The article also discusses safety assessments of the tests as well as modifications required on the EBR-II reactor safety system for conducting required on the EBR-II reactor safety system for the conducting the tests.« less

Coupled reactor kinetics and heat transfer model for heat pipe cooled reactors

NASA Astrophysics Data System (ADS)

Wright, Steven A.; Houts, Michael

2001-02-01

Heat pipes are often proposed as cooling system components for small fission reactors. SAFE-300 and STAR-C are two reactor concepts that use heat pipes as an integral part of the cooling system. Heat pipes have been used in reactors to cool components within radiation tests (Deverall, 1973); however, no reactor has been built or tested that uses heat pipes solely as the primary cooling system. Heat pipe cooled reactors will likely require the development of a test reactor to determine the main differences in operational behavior from forced cooled reactors. The purpose of this paper is to describe the results of a systems code capable of modeling the coupling between the reactor kinetics and heat pipe controlled heat transport. Heat transport in heat pipe reactors is complex and highly system dependent. Nevertheless, in general terms it relies on heat flowing from the fuel pins through the heat pipe, to the heat exchanger, and then ultimately into the power conversion system and heat sink. A system model is described that is capable of modeling coupled reactor kinetics phenomena, heat transfer dynamics within the fuel pins, and the transient behavior of heat pipes (including the melting of the working fluid). This paper focuses primarily on the coupling effects caused by reactor feedback and compares the observations with forced cooled reactors. A number of reactor startup transients have been modeled, and issues such as power peaking, and power-to-flow mismatches, and loading transients were examined, including the possibility of heat flow from the heat exchanger back into the reactor. This system model is envisioned as a tool to be used for screening various heat pipe cooled reactor concepts, for designing and developing test facility requirements, for use in safety evaluations, and for developing test criteria for in-pile and out-of-pile test facilities. .

Liquid-metal flows: Magnetohydrodynamics and applications; Proceedings of the Fifth Beersheba International Seminar on Magnetohydrodynamic Flows and Turbulence, University of the Negev, Beersheba, Israel, Mar. 2-6, 1987

DOE Office of Scientific and Technical Information (OSTI.GOV)

Branover, H.; Mond, M.; Unger, Y.

The present collection of papers on MHD-related uses of liquid metal flows and their applications discusses topics in laminar MHD flows, MHD power generation, metallurgical MHD applications, and two-phase MHD flows. Attention is given to MHD flows with closed streamlines, nonlinear waves in liquid metals under a transverse magnetic field, liquid-metal MHD conversion of nuclear energy to electricity, the testing of optimized MHD conversion (OMACON) systems, and aspects of a liquid-metal induction generator. Also discussed are MHD effects in liquid-metal breeder reactors, a plasma-driven MHD powerplant, modeling the recirculating flows in channel-induction surfaces, the hydrodynamics of aluminum reduction cells, free-surfacemore » determination in a levitation-melting process, the parametric interactions of waves in bubbly liquid metals, and the occurrence of cavitation in water jets.« less

Continuous-flow technology—a tool for the safe manufacturing of active pharmaceutical ingredients.

PubMed

Gutmann, Bernhard; Cantillo, David; Kappe, C Oliver

2015-06-01

In the past few years, continuous-flow reactors with channel dimensions in the micro- or millimeter region have found widespread application in organic synthesis. The characteristic properties of these reactors are their exceptionally fast heat and mass transfer. In microstructured devices of this type, virtually instantaneous mixing can be achieved for all but the fastest reactions. Similarly, the accumulation of heat, formation of hot spots, and dangers of thermal runaways can be prevented. As a result of the small reactor volumes, the overall safety of the process is significantly improved, even when harsh reaction conditions are used. Thus, microreactor technology offers a unique way to perform ultrafast, exothermic reactions, and allows the execution of reactions which proceed via highly unstable or even explosive intermediates. This Review discusses recent literature examples of continuous-flow organic synthesis where hazardous reactions or extreme process windows have been employed, with a focus on applications of relevance to the preparation of pharmaceuticals. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Development of a high-temperature durable catalyst for use in catalytic combustors for advanced automotive gas turbine engines

NASA Astrophysics Data System (ADS)

Tong, H.; Snow, G. C.; Chu, E. K.; Chang, R. L. S.; Angwin, M. J.; Pessagno, S. L.

1981-09-01

Durable catalytic reactors for advanced gas turbine engines were developed. Objectives were: to evaluate furnace aging as a cost effective catalytic reactor screening test, measure reactor degradation as a function of furnace aging, demonstrate 1,000 hours of combustion durability, and define a catalytic reactor system with a high probability of successful integration into an automotive gas turbine engine. Fourteen different catalytic reactor concepts were evaluated, leading to the selection of one for a durability combustion test with diesel fuel for combustion conditions. Eight additional catalytic reactors were evaluated and one of these was successfully combustion tested on propane fuel. This durability reactor used graded cell honeycombs and a combination of noble metal and metal oxide catalysts. The reactor was catalytically active and structurally sound at the end of the durability test.

Development of a high-temperature durable catalyst for use in catalytic combustors for advanced automotive gas turbine engines

NASA Technical Reports Server (NTRS)

Tong, H.; Snow, G. C.; Chu, E. K.; Chang, R. L. S.; Angwin, M. J.; Pessagno, S. L.

1981-01-01

Durable catalytic reactors for advanced gas turbine engines were developed. Objectives were: to evaluate furnace aging as a cost effective catalytic reactor screening test, measure reactor degradation as a function of furnace aging, demonstrate 1,000 hours of combustion durability, and define a catalytic reactor system with a high probability of successful integration into an automotive gas turbine engine. Fourteen different catalytic reactor concepts were evaluated, leading to the selection of one for a durability combustion test with diesel fuel for combustion conditions. Eight additional catalytic reactors were evaluated and one of these was successfully combustion tested on propane fuel. This durability reactor used graded cell honeycombs and a combination of noble metal and metal oxide catalysts. The reactor was catalytically active and structurally sound at the end of the durability test.

Exploratory development of a glass ceramic automobile thermal reactor. [anti-pollution devices

NASA Technical Reports Server (NTRS)

Gould, R. E.; Petticrew, R. W.

1973-01-01

This report summarizes the design, fabrication and test results obtained for glass-ceramic (CER-VIT) automotive thermal reactors. Several reactor designs were evaluated using both engine-dynamometer and vehicle road tests. A maximum reactor life of about 330 hours was achieved in engine-dynamometer tests with peak gas temperatures of about 1065 C (1950 F). Reactor failures were mechanically induced. No evidence of chemical degradation was observed. It was concluded that to be useful for longer times, the CER-VIT parts would require a mounting system that was an improvement over those tested in this program. A reactor employing such a system was designed and fabricated.

Workshop on development of radionuclide getters for the Yucca Mountain waste repository: proceedings.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moore, Robert Charles; Lukens, Wayne W.

The proposed Yucca Mountain repository, located in southern Nevada, is to be the first facility for permanent disposal of spent reactor fuel and high-level radioactive waste in the United States. Total Systems Performance Assessment (TSPA) analysis has indicated that among the major radionuclides contributing to dose are technetium, iodine, and neptunium, all of which are highly mobile in the environment. Containment of these radionuclides within the repository is a priority for the Yucca Mountain Project (YMP). These proceedings review current research and technology efforts for sequestration of the radionuclides with a focus on technetium, iodine, and neptunium. This workshop alsomore » covered issues concerning the Yucca Mountain environment and getter characteristics required for potential placement into the repository.« less

Commercial-scale demonstration of the Liquid Phase Methanol (LPMEOH{trademark}) process. Technical progress report number 11, January 1--March 31, 1997

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1997-06-11

During this quarter, the third draft of the Topical Report on Process Economics Studies was issued for review. A recommendation to continue with design verification testing on the coproduction of methanol and dimethyl ether (DME) was made. A liquid phase dimethyl ether (LPDME) catalyst system with reasonable long-term activity and stability is being developed, and a decision to proceed with a proof-of-concept test run at the LaPorte Alternative Fuels Development Unit (AFDU) is pending the release of a memo from Air Products on the catalyst targets and corresponding economics for a commercially successful LPDME catalyst. The off-site product-use test planmore » is to be updated in June of 1997. During this quarter, Air Products and Acurex Environmental Corporation continued developing the listing of product-use test participants who are involved in fuel cell, transportation, and stationary power plant applications. Start-up activities (Task 3.1) began during the reporting period, and coal-derived synthesis gas (syngas) was introduced to the demonstration unit. The recycle compressor was tested successfully on syngas at line pressure of 700 psig, and the reactor loop reached 220 C for carbonyl burnout. Iron carbonyl in the balanced gas feed remained below the 10 ppbv detection limit for all samples but one. Within the reactor loop, iron carbonyl levels peaked out near 200 ppbv after about 40 hours on-stream, before decreasing to between 10--20 ppbv at 160 hours on -stream. Nickel carbonyl measurements reached a peak of about 60 ppbv, and decreased at all sampling locations to below the 10 ppbv detection limit by 70 hours on-stream. Catalyst activation of the nine 2,250 lb batches required for the initial catalyst charge began and concluded. All batches met or slightly exceeded the theoretical maximum uptake of 2.82 SCF of reducing gas/lb catalyst.« less

ADVANCED REACTIVITY MEASUREMENT FACILITY, TRA660, INTERIOR. REACTOR INSIDE TANK. METAL ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ADVANCED REACTIVITY MEASUREMENT FACILITY, TRA-660, INTERIOR. REACTOR INSIDE TANK. METAL WORK PLATFORM ABOVE. THE REACTOR WAS IN A SMALL WATER-FILLED POOL. INL NEGATIVE NO. 66-6373. Unknown Photographer, ca. 1966 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

MATERIALS TESTING REACTOR (MTR) BUILDING, TRA603. CONTEXTUAL VIEW OF MTR ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

MATERIALS TESTING REACTOR (MTR) BUILDING, TRA-603. CONTEXTUAL VIEW OF MTR BUILDING SHOWING NORTH SIDES OF THE HIGH-BAY REACTOR BUILDING, ITS SECOND/THIRD FLOOR BALCONY LEVEL, AND THE ATTACHED ONE-STORY OFFICE/LABORATORY BUILDING, TRA-604. CAMERA FACING SOUTHEAST. VERTICAL CONCRETE-SHROUDED BEAMS SUPPORT PRECAST CONCRETE PANELS. CONCRETE PROJECTION FORMED AS A BUNKER AT LEFT OF VIEW IS TRA-657, PLUG STORAGE BUILDING. INL NEGATIVE NO. HD46-42-1. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Lessons Learned about Liquid Metal Reactors from FFTF Experience

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wootan, David W.; Casella, Andrew M.; Omberg, Ronald P.

2016-09-20

The Fast Flux Test Facility (FFTF) is the most recent liquid-metal reactor (LMR) to operate in the United States, from 1982 to 1992. FFTF is located on the DOE Hanford Site near Richland, Washington. The 400-MWt sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission test reactor was designed specifically to irradiate Liquid Metal Fast Breeder Reactor (LMFBR) fuel and components in prototypical temperature and flux conditions. FFTF played a key role in LMFBR development and testing activities. The reactor provided extensive capability for in-core irradiation testing, including eight core positions that could be used with independent instrumentation for the test specimens.more » In addition to irradiation testing capabilities, FFTF provided long-term testing and evaluation of plant components and systems for LMFBRs. The FFTF was highly successful and demonstrated outstanding performance during its nearly 10 years of operation. The technology employed in designing and constructing this reactor, as well as information obtained from tests conducted during its operation, can significantly influence the development of new advanced reactor designs in the areas of plant system and component design, component fabrication, fuel design and performance, prototype testing, site construction, and reactor operations. The FFTF complex included the reactor, as well as equipment and structures for heat removal, containment, core component handling and examination, instrumentation and control, and for supplying utilities and other essential services. The FFTF Plant was designed using a “system” concept. All drawings, specifications and other engineering documentation were organized by these systems. Efforts have been made to preserve important lessons learned during the nearly 10 years of reactor operation. A brief summary of Lessons Learned in the following areas will be discussed: Acceptance and Startup Testing of FFTF FFTF Cycle Reports« less

WORKER STACKS GRAPHITE BLOCKS AGAINST INNER SOUTH WALL OF REACTOR. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

WORKER STACKS GRAPHITE BLOCKS AGAINST INNER SOUTH WALL OF REACTOR. INL NEGATIVE NO. 3925. Unknown Photographer, 12/14/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Eddy Current Flow Measurements in the FFTF

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nielsen, Deborah L.; Polzin, David L.; Omberg, Ronald P.

2017-02-02

The Fast Flux Test Facility (FFTF) is the most recent liquid metal reactor (LMR) to be designed, constructed, and operated by the U.S. Department of Energy (DOE). The 400-MWt sodium-cooled, fast-neutron flux reactor plant was designed for irradiation testing of nuclear reactor fuels and materials for liquid metal fast breeder reactors. Following shut down of the Clinch River Breeder Reactor Plant (CRBRP) project in 1983, FFTF continued to play a key role in providing a test bed for demonstrating performance of advanced fuel designs and demonstrating operation, maintenance, and safety of advanced liquid metal reactors. The FFTF Program provides valuablemore » information for potential follow-on reactor projects in the areas of plant system and component design, component fabrication, fuel design and performance, prototype testing, site construction, and reactor control and operations. This report provides HEDL-TC-1344, “ECFM Flow Measurements in the FFTF Using Phase-Sensitive Detectors”, March 1979.« less

Review of Transient Testing of Fast Reactor Fuels in the Transient REActor Test Facility (TREAT)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jensen, C.; Wachs, D.; Carmack, J.

The restart of the Transient REActor Test (TREAT) facility provides a unique opportunity to engage the fast reactor fuels community to reinitiate in-pile experimental safety studies. Historically, the TREAT facility played a critical role in characterizing the behavior of both metal and oxide fast reactor fuels under off-normal conditions, irradiating hundreds of fuel pins to support fast reactor fuel development programs. The resulting test data has provided validation for a multitude of fuel performance and severe accident analysis computer codes. This paper will provide a review of the historical database of TREAT experiments including experiment design, instrumentation, test objectives, andmore » salient findings. Additionally, the paper will provide an introduction to the current and future experiment plans of the U.S. transient testing program at TREAT.« less

ENGINEERING TEST REACTOR (ETR) BUILDING, TRA642. CONTEXTUAL VIEW, CAMERA FACING ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ENGINEERING TEST REACTOR (ETR) BUILDING, TRA-642. CONTEXTUAL VIEW, CAMERA FACING EAST. VERTICAL METAL SIDING. ROOF IS SLIGHTLY ELEVATED AT CENTER LINE FOR DRAINAGE. WEST SIDE OF ETR COMPRESSOR BUILDING, TRA-643, PROJECTS TOWARD LEFT AT FAR END OF ETR BUILDING. INL NEGATIVE NO. HD46-37-1. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

28 CFR 28.24 - Exceptions based on the results of judicial proceedings.

Code of Federal Regulations, 2010 CFR

2010-07-01

... judicial proceedings. 28.24 Section 28.24 Judicial Administration DEPARTMENT OF JUSTICE DNA IDENTIFICATION... inapplicable in two circumstances relating to the results of judicial proceedings: (a) Judicial denial of DNA testing. Section 3600A(c)(1) exempts situations in which a court has denied a motion for DNA testing under...

28 CFR 28.24 - Exceptions based on the results of judicial proceedings.

Code of Federal Regulations, 2011 CFR

2011-07-01

... judicial proceedings. 28.24 Section 28.24 Judicial Administration DEPARTMENT OF JUSTICE DNA IDENTIFICATION... inapplicable in two circumstances relating to the results of judicial proceedings: (a) Judicial denial of DNA testing. Section 3600A(c)(1) exempts situations in which a court has denied a motion for DNA testing under...

28 CFR 28.24 - Exceptions based on the results of judicial proceedings.

Code of Federal Regulations, 2013 CFR

2013-07-01

... judicial proceedings. 28.24 Section 28.24 Judicial Administration DEPARTMENT OF JUSTICE DNA IDENTIFICATION... inapplicable in two circumstances relating to the results of judicial proceedings: (a) Judicial denial of DNA testing. Section 3600A(c)(1) exempts situations in which a court has denied a motion for DNA testing under...

28 CFR 28.24 - Exceptions based on the results of judicial proceedings.

Code of Federal Regulations, 2012 CFR

2012-07-01

... judicial proceedings. 28.24 Section 28.24 Judicial Administration DEPARTMENT OF JUSTICE DNA IDENTIFICATION... inapplicable in two circumstances relating to the results of judicial proceedings: (a) Judicial denial of DNA testing. Section 3600A(c)(1) exempts situations in which a court has denied a motion for DNA testing under...

28 CFR 28.24 - Exceptions based on the results of judicial proceedings.

Code of Federal Regulations, 2014 CFR

2014-07-01

... judicial proceedings. 28.24 Section 28.24 Judicial Administration DEPARTMENT OF JUSTICE DNA IDENTIFICATION... inapplicable in two circumstances relating to the results of judicial proceedings: (a) Judicial denial of DNA testing. Section 3600A(c)(1) exempts situations in which a court has denied a motion for DNA testing under...

Space reactor fuel element testing in upgraded TREAT

NASA Astrophysics Data System (ADS)

Todosow, M.; Bezler, P.; Ludewig, H.; Kato, W. Y.

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc.; a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR); NERVA-derivative; and other concepts are discussed. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggest that full-scale PBR elements could be tested at an average energy deposition of approximately 60-80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of approximately 100 MW/L may be achievable.

Space reactor fuel element testing in upgraded TREAT

NASA Astrophysics Data System (ADS)

Todosow, Michael; Bezler, Paul; Ludewig, Hans; Kato, Walter Y.

1993-01-01

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., is a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggests that full-scale PBR elements could be tested at an average energy deposition of ˜60-80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperture limit, average energy deposition of ˜100 MW/L may be achievable.

Proceedings of the international meeting on thermal nuclear reactor safety. Vol. 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

Separate abstracts are included for each of the papers presented concerning current issues in nuclear power plant safety; national programs in nuclear power plant safety; radiological source terms; probabilistic risk assessment methods and techniques; non LOCA and small-break-LOCA transients; safety goals; pressurized thermal shocks; applications of reliability and risk methods to probabilistic risk assessment; human factors and man-machine interface; and data bases and special applications.

Design optimization of first wall and breeder unit module size for the Indian HCCB blanket module

NASA Astrophysics Data System (ADS)

Deepak, SHARMA; Paritosh, CHAUDHURI

2018-04-01

The Indian test blanket module (TBM) program in ITER is one of the major steps in the Indian fusion reactor program for carrying out the R&D activities in the critical areas like design of tritium breeding blankets relevant to future Indian fusion devices (ITER relevant and DEMO). The Indian Lead–Lithium Cooled Ceramic Breeder (LLCB) blanket concept is one of the Indian DEMO relevant TBM, to be tested in ITER as a part of the TBM program. Helium-Cooled Ceramic Breeder (HCCB) is an alternative blanket concept that consists of lithium titanate (Li2TiO3) as ceramic breeder (CB) material in the form of packed pebble beds and beryllium as the neutron multiplier. Specifically, attentions are given to the optimization of first wall coolant channel design and size of breeder unit module considering coolant pressure and thermal loads for the proposed Indian HCCB blanket based on ITER relevant TBM and loading conditions. These analyses will help proceeding further in designing blankets for loads relevant to the future fusion device.

SPERT Destructive Test - I on Aluminum, Highly Enriched Plate Type Core

ScienceCinema

None

2018-01-16

SPERT - Special Power Excursion Reactor Tests Destructive Test number 1 On Aluminum, Highly Enriched Plate Type Core. A test studying the behavior of the reactor under destructive conditions on a light water moderated pool-type reactor with a plate-type core.

Design and evaluation of experimental ceramic automobile thermal reactors

NASA Technical Reports Server (NTRS)

Stone, P. L.; Blankenship, C. P.

1974-01-01

The paper summarizes the results obtained in an exploratory evaluation of ceramics for automobile thermal reactors. Candidate ceramic materials were evaluated in several reactor designs using both engine dynamometer and vehicle road tests. Silicon carbide contained in a corrugated metal support structure exhibited the best performance, lasting 1100 hours in engine dynamometer tests and for more than 38,600 kilimeters (24,000 miles) in vehicle road tests. Although reactors containing glass-ceramic components did not perform as well as silicon carbide, the glass-ceramics still offer good potential for reactor use with improved reactor designs.

Design and evaluation of experimental ceramic automobile thermal reactors

NASA Technical Reports Server (NTRS)

Stone, P. L.; Blankenship, C. P.

1974-01-01

The results obtained in an exploratory evaluation of ceramics for automobile thermal reactors are summarized. Candidate ceramic materials were evaluated in several reactor designs by using both engine-dynamometer and vehicle road tests. Silicon carbide contained in a corrugated-metal support structure exhibited the best performance, lasting 1100 hr in engine-dynamometer tests and more than 38,600 km (24000 miles) in vehicle road tests. Although reactors containing glass-ceramic components did not perform as well as those containing silicon carbide, the glass-ceramics still offer good potential for reactor use with improved reactor designs.

77 FR 63897 - Notice of License Terminations for National Aeronautics and Space Administration; Plum Brook...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-17

... test reactor, constructed to perform irradiation testing of fueled and unfueled experiments for space... constructed to test ``mock-up'' irradiation components for the Plum Brook Reactor. The reactors operated from...

Screening study for evaluation of the potential for system 80+ to consume excess plutonium - Volume 1. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1994-04-30

As part of the U.S. effort to evaluate technologies offering solutions for the safe disposal or utilization of surplus nuclear materials, the fiscal year 1993 Energy and Water Appropriations legislation provided the Department of Energy (DOE) the necessary funds to conduct multi-phased studies to determine the technical feasibility of using reactor technologies for the triple mission of burning weapons grade plutonium, producing tritium for the existing smaller weapons stockpile, and generating commercial electricity. DOE limited the studies to five advanced reactor designs. Among the technologies selected is the ABB-Combustion Engineering (ABB-CE) System 80+. The DOE study, currently in Phase ID,more » is proceeding with a more detailed evaluation of the design`s capability for plutonium disposition.« less

A document review to characterize Atomic International SNAP fuels shipped to INEL 1966--1973

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wahnschaffe, S.D.; Lords, R.E.; Kneff, D.W.

1995-09-01

This report provides the results of a document search and review study to obtain information on the spent fuels for the following six Nuclear Auxiliary Power (SNAP) reactor cores now stored at the Idaho National Engineering Laboratory (INEL): SNAP-2 Experimental Reactor, SNAP-2 Development Reactor, SNAP-10A Ground Test Reactor, SNAP-8 Experimental Reactor, SNAP-8 Development Reactor, and Shield Test Reactor. The report also covers documentation on SNAP fuel materials from four in-pile materials tests: NAA-82-1, NAA-115-2, NAA-117-1, and NAA-121. Pieces of these fuel materials are also stored at INEL as part of the SNAP fuel shipments.

Analytical design and performance studies of nuclear furnace tests of small nuclear light bulb models

NASA Technical Reports Server (NTRS)

Latham, T. S.; Rodgers, R. J.

1972-01-01

Analytical studies were continued to identify the design and performance characteristics of a small-scale model of a nuclear light bulb unit cell suitable for testing in a nuclear furnace reactor. Emphasis was placed on calculating performance characteristics based on detailed radiant heat transfer analyses, on designing the test assembly for ease of insertion, connection, and withdrawal at the reactor test cell, and on determining instrumentation and test effluent handling requirements. In addition, a review of candidate test reactors for future nuclear light bulb in-reactor tests was conducted.

Operators in the Plum Brook Reactor Facility Control Room

NASA Image and Video Library

1970-03-21

Donald Rhodes, left, and Clyde Greer, right, monitor the operation of the National Aeronautics and Space Administration’s (NASA) Plum Brook Reactor Facility from the control room. The 60-megawatt test reactor, NASA’s only reactor, was the eighth largest test reactor in the world. The facility was built by the Lewis Research Center in the late 1950s to study the effects of radiation on different materials that could be used to construct nuclear propulsion systems for aircraft or rockets. The reactor went critical for the first time in 1961. For the next two years, two operators were on duty 24 hours per day working on the fission process until the reactor reached its full-power level in 1963. Reactor Operators were responsible for monitoring and controlling the reactor systems. Once the reactor was running under normal operating conditions, the work was relatively uneventful. Normally the reactor was kept at a designated power level within certain limits. Occasionally the operators had to increase the power for a certain test. The shift supervisor and several different people would get together and discuss the change before boosting the power. All operators were required to maintain a Reactor Operator License from the Atomic Energy Commission. The license included six months of training, an eight-hour written exam, a four-hour walkaround, and testing on the reactor controls.

26 CFR 1.103-8 - Interest on bonds to finance certain exempt facilities.

Code of Federal Regulations, 2014 CFR

2014-04-01

... proceeds of such issue is to be used to provide one or more of the exempt facilities listed in... proceeds are so used. For purposes of this “substantially all” test, two rules apply. First, proceeds are... becomes applicable where the bond issue meets both the trade or business and the security interest tests...

26 CFR 1.103-8 - Interest on bonds to finance certain exempt facilities.

Code of Federal Regulations, 2010 CFR

2010-04-01

... proceeds of such issue is to be used to provide one or more of the exempt facilities listed in... proceeds are so used. For purposes of this “substantially all” test, two rules apply. First, proceeds are... becomes applicable where the bond issue meets both the trade or business and the security interest tests...

26 CFR 1.103-8 - Interest on bonds to finance certain exempt facilities.

Code of Federal Regulations, 2012 CFR

2012-04-01

... proceeds of such issue is to be used to provide one or more of the exempt facilities listed in... proceeds are so used. For purposes of this “substantially all” test, two rules apply. First, proceeds are... becomes applicable where the bond issue meets both the trade or business and the security interest tests...

26 CFR 1.103-8 - Interest on bonds to finance certain exempt facilities.

Code of Federal Regulations, 2013 CFR

2013-04-01

... proceeds of such issue is to be used to provide one or more of the exempt facilities listed in... proceeds are so used. For purposes of this “substantially all” test, two rules apply. First, proceeds are... becomes applicable where the bond issue meets both the trade or business and the security interest tests...

26 CFR 1.103-8 - Interest on bonds to finance certain exempt facilities.

Code of Federal Regulations, 2011 CFR

2011-04-01

... proceeds of such issue is to be used to provide one or more of the exempt facilities listed in... proceeds are so used. For purposes of this “substantially all” test, two rules apply. First, proceeds are... becomes applicable where the bond issue meets both the trade or business and the security interest tests...

Space reactor fuel element testing in upgraded TREAT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Todosow, M.; Bezler, P.; Ludewig, H.

1993-01-14

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. initial results suggest that full-scale PBR, elements could be tested at an average energy deposition of {approximately}60--80 MW-s/L in the current TREAT reactor. Ifmore » the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of {approximately}100 MW/L may be achievable.« less

Space reactor fuel element testing in upgraded TREAT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Todosow, M.; Bezler, P.; Ludewig, H.

1993-05-01

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. initial results suggest that full-scale PBR, elements could be tested at an average energy deposition of {approximately}60--80 MW-s/L in the current TREAT reactor. Ifmore » the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of {approximately}100 MW/L may be achievable.« less

Oxidation of aluminum alloy cladding for research and test reactor fuel

NASA Astrophysics Data System (ADS)

Kim, Yeon Soo; Hofman, G. L.; Robinson, A. B.; Snelgrove, J. L.; Hanan, N.

2008-08-01

The oxide thicknesses on aluminum alloy cladding were measured for the test plates from irradiation tests RERTR-6 and 7A in the ATR (advanced test reactor). The measured thicknesses were substantially lower than those of test plates with similar power from other reactors available in the literature. The main reason is believed to be due to the lower pH (pH 5.1-5.3) of the primary coolant water in the ATR than in the other reactors (pH 5.9-6.5) for which we have data. An empirical model for oxide film thickness predictions on aluminum alloy used as fuel cladding in the test reactors was developed as a function of irradiation time, temperature, surface heat flux, pH, and coolant flow rate. The applicable ranges of pH and coolant flow rates cover most research and test reactors. The predictions by the new model are in good agreement with the in-pile test data available in the literature as well as with the RERTR test data measured in the ATR.

SNAP 10A FS-3 reactor performance

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hawley, J.P.; Johnson, R.A.

1966-08-15

SNAP 10FS-3 was the first flight-qualified SNAP reactor system to be operated in a simulated space environment. Prestart-up qualification testing, automatic start-up, endurance period performance, extended operation test and reactor shutdown are described as they affected, or were affected by, overall reactor performance. Performance of the reactor control system and the diagnostic instrumentation is critically evaluted.

Development of RF plasma simulations of in-reactor tests of small models of the nuclear light bulb fuel region

NASA Technical Reports Server (NTRS)

Roman, W. C.; Jaminet, J. F.

1972-01-01

Experiments were conducted to develop test configurations and technology necessary to simulate the thermal environment and fuel region expected to exist in in-reactor tests of small models of nuclear light bulb configurations. Particular emphasis was directed at rf plasma tests of approximately full-scale models of an in-reactor cell suitable for tests in Los Alamos Scientific Laboratory's Nuclear Furnace. The in-reactor tests will involve vortex-stabilized fissioning uranium plasmas of approximately 200-kW power, 500-atm pressure and equivalent black-body radiating temperatures between 3220 and 3510 K.

Exploratory evaluation of ceramics for automobile thermal reactors

NASA Technical Reports Server (NTRS)

Stone, P. L.; Blankenship, C. P.

1972-01-01

An exploratory evaluation of ceramics for automobile thermal reactors was conducted. Potential ceramic materials were evaluated in several reactor designs using both engine dynamometer and vehicle road tests. Silicon carbide contained in a corrugated metal support structure exhibited the best performance lasting over 800 hours in engine dynamometer tests and over 15,000 miles (24,200 km) of vehicle road tests. Reactors containing glass-ceramic components did not perform as well as silicon carbide. But the glass-ceramics still offer good potential for reactor use. The results of this study are considered to be a reasonable demonstration of the potential use of ceramics in thermal reactors.

Cavity temperature and flow characteristics in a gas-core test reactor

NASA Technical Reports Server (NTRS)

Putre, H. A.

1973-01-01

A test reactor concept for conducting basic studies on a fissioning uranium plasma and for testing various gas-core reactor concepts is analyzed. The test reactor consists of a conventional fuel-element region surrounding a 61-cm-(2-ft-) diameter cavity region which contains the plasma experiment. The fuel elements provide the neutron flux for the cavity region. The design operating conditions include 60-MW reactor power, 2.7-MW cavity power, 200-atm cavity pressure, and an average uranium plasma temperature of 15,000 K. The analytical results are given for cavity radiant heat transfer, hydrogen transpiration cooling, and uranium wire or powder injection.

PBF Reactor Building (PER620). PBF crane holds fuel test assembly ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PBF Reactor Building (PER-620). PBF crane holds fuel test assembly aloft prior to lowering into reactor for test. Date: 1982. INEEL negative no. 82-4909 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

77 FR 16868 - Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-03-22

... Fuel Elements for Use in Research and Test Reactors AGENCY: Nuclear Regulatory Commission. ACTION... Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test Reactors.'' This guide describes... plate-type uranium-aluminum fuel elements used in research and test reactors (RTRs). DATES: Submit...

Dynamic Response Testing in an Electrically Heated Reactor Test Facility

NASA Astrophysics Data System (ADS)

Bragg-Sitton, Shannon M.; Morton, T. J.

2006-01-01

Non-nuclear testing can be a valuable tool in the development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and fueled nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe (HP) cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system. Reactivity feedback calculations were then based on a bulk reactivity feedback coefficient and measured average core temperature. This paper presents preliminary results from similar dynamic testing of a direct drive gas cooled reactor system (DDG), demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. Although the HP and DDG designs both utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility. Planned system upgrades to allow implementation of higher fidelity dynamic testing are also discussed. Proposed DDG testing will utilize a higher fidelity point kinetics model to control core power transients, and reactivity feedback will be based on localized feedback coefficients and several independent temperature measurements taken within the core block. This paper presents preliminary test results and discusses the methodology that will be implemented in follow-on DDG testing and the additional instrumentation required to implement high fidelity dynamic testing.

Preliminary plan for testing a thermionic reactor in the Plum Brook Space Power Facility

NASA Technical Reports Server (NTRS)

Haley, F. A.

1972-01-01

A preliminary plan is presented for testing a thermionic reactor in the Plum Brook Space Power Facility (SPF). A technical approach, cost estimate, manpower estimate, and schedule are presented to cover a 2 year full power reactor test.

Design and Status of RERTR Irradiation Tests in the Advanced Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Daniel M. Wachs; Richard G. Ambrosek; Gray Chang

2006-10-01

Irradiation testing of U-Mo based fuels is the central component of the Reduced Enrichment for Research and Test Reactors (RERTR) program fuel qualification plan. Several RERTR tests have recently been completed or are planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory in Idaho Falls, ID. Four mini-plate experiments in various stages of completion are described in detail, including the irradiation test design, objectives, and irradiation conditions. Observations made during and after the in-reactor RERTR-7A experiment breach are summarized. The irradiation experiment design and planned irradiation conditions for full-size plate test are described. Progressmore » toward element testing will be reviewed.« less

Preliminary Options Assessment of Versatile Irradiation Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sen, Ramazan Sonat

The objective of this report is to summarize the work undertaken at INL from April 2016 to January 2017 and aimed at analyzing some options for designing and building a versatile test reactor; the scope of work was agreed upon with DOE-NE. Section 2 presents some results related to KNK II and PRISM Mod A. Section 3 presents some alternatives to the VCTR presented in [ ] as well as a neutronic parametric study to assess the minimum power requirement needed for a 235U metal fueled fast test reactor capable to generate a fast (>100 keV) flux of 4.0 xmore » 1015 n /cm2-s at the test location. Section 4 presents some results regarding a fundamental characteristic of test reactors, namely displacement per atom (dpa) in test samples. Section 5 presents the INL assessment of the ANL fast test reactor design FASTER. Section 6 presents a summary.« less

High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sterbentz, James William; Bayless, Paul David; Nelson, Lee Orville

2016-01-01

A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sterbentz, James William; Bayless, Paul David; Nelson, Lee Orville

2016-03-01

A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

Space station prototype Sabatier reactor design verification testing

NASA Technical Reports Server (NTRS)

Cusick, R. J.

1974-01-01

A six-man, flight prototype carbon dioxide reduction subsystem for the SSP ETC/LSS (Space Station Prototype Environmental/Thermal Control and Life Support System) was developed and fabricated for the NASA-Johnson Space Center between February 1971 and October 1973. Component design verification testing was conducted on the Sabatier reactor covering design and off-design conditions as part of this development program. The reactor was designed to convert a minimum of 98 per cent hydrogen to water and methane for both six-man and two-man reactant flow conditions. Important design features of the reactor and test conditions are described. Reactor test results are presented that show design goals were achieved and off-design performance was stable.

Modeling an unmitigated thermal quench event in a large field magnet in a DEMO reactor

DOE PAGES

Merrill, Brad J.

2015-03-25

The superconducting magnet systems of future fusion reactors, such as a Demonstration Power Plant (DEMO), will produce magnetic field energies in the 10 s of GJ range. The release of this energy during a fault condition could produce arcs that can damage the magnets of these systems. The public safety consequences of such events must be explored for a DEMO reactor because the magnets are located near the DEMO's primary radioactive confinement barrier, the reactor's vacuum vessel (VV). Great care will be taken in the design of DEMO's magnet systems to detect and provide a rapid field energy dump tomore » avoid any accidents conditions. During an event when a fault condition proceeds undetected, the potential of producing melting of the magnet exists. If molten material from the magnet impinges on the walls of the VV, these walls could fail, resulting in a pathway for release of radioactive material from the VV. A model is under development at Idaho National Laboratory (INL) called MAGARC to investigate the consequences of this accident in a large toroidal field (TF) coil. Recent improvements to this model are described in this paper, along with predictions for a DEMO relevant event in a toroidal field magnet.« less

Recent Results from the Daya Bay Reactor Neutrino Experiment

NASA Astrophysics Data System (ADS)

Huang, En-Chuan

2016-11-01

The Daya Bay Reactor Neutrino Experiment is designed to precisely measure the mixing parameter sin2 2θ13 via relative measurements with eight functionally identical antineutrino detectors (ADs). In 2012, Daya Bay has first measured a non-zero sin2 2θ13 value with a significance larger than 5σ with the first six ADs. With the installation of two new ADs to complete the full configuration, Daya Bay has continued to increase statistics and lower systematic uncertainties for better precision of sin2 2θ13 and for the exploration of other physics topics. In this proceeding, the latest analysis results of sin2 2θ13 and |Δm 2 ee|, including a measurement made with neutron capture on Gadolinium and an independent measurement made with neutron capture on hydrogen are presented. The latest results of the search for sterile neutrino in the mass splitting range of 10-3 eV2 < |Δm 2 41| < 0.3 eV2 and the absolute measurement of the rate and energy spectrum of reactor antineutrinos will also be presented.

Study for requirement of advanced long life small modular fast reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tak, Taewoo, E-mail: ttwispy@unist.ac.kr; Choe, Jiwon, E-mail: chi91023@unist.ac.kr; Jeong, Yongjin, E-mail: yjjeong09@unist.ac.kr

2016-01-22

To develop an advanced long-life SMR core concept, the feasibility of the long-life breed-and-burn core concept has been assessed and the preliminary selection on the reactor design requirement such as fuel form, coolant material has been performed. With the simplified cigar-type geometry of 8m-tall CANDLE reactor concept, it has demonstrated the strengths of breed-and-burn strategy. There is a saturation region in the graph for the multiplication factors, which means that a steady breeding is being proceeded along the axial direction. The propagation behavior of the CANDLE core can be also confirmed through the evolution of the axial power profile. Coolantmore » material is expected to have low melting point, density, viscosity and absorption cross section and a high boiling point, specific heat, and thermal conductivity. In this respect, sodium is preferable material for a coolant of this nuclear power plant system. The metallic fuel has harder spectrum compared to the oxide and carbide fuel, which is favorable to increase the breeding and extend the cycle length.« less

Study for requirement of advanced long life small modular fast reactor

NASA Astrophysics Data System (ADS)

Tak, Taewoo; Choe, Jiwon; Jeong, Yongjin; Lee, Deokjung; Kim, T. K.

2016-01-01

To develop an advanced long-life SMR core concept, the feasibility of the long-life breed-and-burn core concept has been assessed and the preliminary selection on the reactor design requirement such as fuel form, coolant material has been performed. With the simplified cigar-type geometry of 8m-tall CANDLE reactor concept, it has demonstrated the strengths of breed-and-burn strategy. There is a saturation region in the graph for the multiplication factors, which means that a steady breeding is being proceeded along the axial direction. The propagation behavior of the CANDLE core can be also confirmed through the evolution of the axial power profile. Coolant material is expected to have low melting point, density, viscosity and absorption cross section and a high boiling point, specific heat, and thermal conductivity. In this respect, sodium is preferable material for a coolant of this nuclear power plant system. The metallic fuel has harder spectrum compared to the oxide and carbide fuel, which is favorable to increase the breeding and extend the cycle length.

CRITICAL EXPERIMENT TANK (CET) REACTOR HAZARDS SUMMARY

DOE Office of Scientific and Technical Information (OSTI.GOV)

Becar, N.J.; Kunze, J.F.; Pincock, G..D.

1961-03-31

The Critical Experiment Tank (CET) reactor assembly, the associated systems, and the Low Power Test Facility in which the reactor is to be operated are described. An evaluation and summary of the hazards associated with the operation of the CET reactor in the LPTF at the ldsho Test Station are also presented. (auth)

ETRCF, TRA654, INTERIOR. REACTOR OPERATED IN WATERFILLED TANK. CAMERA LOOKS ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR-CF, TRA-654, INTERIOR. REACTOR OPERATED IN WATER-FILLED TANK. CAMERA LOOKS DOWN FROM ABOVE UPON LATER (NON-NUCLEAR) EXPERIMENTAL GEAR. INL NEGATIVE NO. HD24-1-1. Mike Crane, Photographer, ca. 2003 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

MTR, SOUTH FACE OF REACTOR. SPECIAL SUPPLEMENTAL SHIELDING WAS REQUIRED ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

MTR, SOUTH FACE OF REACTOR. SPECIAL SUPPLEMENTAL SHIELDING WAS REQUIRED OUTSIDE OF MTR FOR EXPERIMENTS. THE AIRCRAFT NUCLEAR PROPULSION PROJECT DOMINATED THE USE OF THIS PART OF THE MTR. INL NEGATIVE NO. 7225. Unknown Photographer, 11/28/1952 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Nuclear Science Symposium, 21st, Scintillation and Semiconductor Counter Symposium, 14th, and Nuclear Power Systems Symposium, 6th, Washington, D.C., December 11-13, 1974, Proceedings

NASA Technical Reports Server (NTRS)

1975-01-01

Papers are presented dealing with latest advances in the design of scintillation counters, semiconductor radiation detectors, gas and position sensitive radiation detectors, and the application of these detectors in biomedicine, satellite instrumentation, and environmental and reactor instrumentation. Some of the topics covered include entopistic scintillators, neutron spectrometry by diamond detector for nuclear radiation, the spherical drift chamber for X-ray imaging applications, CdTe detectors in radioimmunoassay analysis, CAMAC and NIM systems in the space program, a closed loop threshold calibrator for pulse height discriminators, an oriented graphite X-ray diffraction telescope, design of a continuous digital-output environmental radon monitor, and the optimization of nanosecond fission ion chambers for reactor physics. Individual items are announced in this issue.

Core Dynamics Analysis for Reactivity Insertion and Loss of Coolant Flow Tests Using the High Temperature Engineering Test Reactor

NASA Astrophysics Data System (ADS)

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Takeda, Tetsuaki

Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are in progress to verify its inherent safety features and improve the safety technology and design methodology for High-temperature Gas-cooled Reactors (HTGRs). The reactivity insertion test is one of the safety demonstration tests for the HTTR. This test simulates the rapid increase in the reactor power by withdrawing the control rod without operating the reactor power control system. In addition, the loss of coolant flow tests has been conducted to simulate the rapid decrease in the reactor power by tripping one, two or all out of three gas circulators. The experimental results have revealed the inherent safety features of HTGRs, such as the negative reactivity feedback effect. The numerical analysis code, which was named-ACCORD-, was developed to analyze the reactor dynamics including the flow behavior in the HTTR core. We have modified this code to use a model with four parallel channels and twenty temperature coefficients. Furthermore, we added another analytical model of the core for calculating the heat conduction between the fuel channels and the core in the case of the loss of coolant flow tests. This paper describes the validation results for the newly developed code using the experimental results. Moreover, the effect of the model is formulated quantitatively with our proposed equation. Finally, the pre-analytical result of the loss of coolant flow test by tripping all gas circulators is also discussed.

JPRS Report, Science & Technology, China: Energy.

DTIC Science & Technology

1992-03-30

breeder reactors should become...the primary type of reactors . In developing breeder reactors , we should follow the path of using metal fuel. Breeder reactors give us more time to...first reactor used for power generation was a fast reactor : the " Breeder 1" reactor at the Idaho National Reactor Test Center which was used to

HEDL FACILITIES CATALOG 400 AREA

DOE Office of Scientific and Technical Information (OSTI.GOV)

MAYANCSIK BA

1987-03-01

The purpose of this project is to provide a sodium-cooled fast flux test reactor designed specifically for irradiation testing of fuels and materials and for long-term testing and evaluation of plant components and systems for the Liquid Metal Reactor (LMR) Program. The FFTF includes the reactor, heat removal equipment and structures, containment, core component handling and examination, instrumentation and control, and utilities and other essential services. The complex array of buildings and equipment are arranged around the Reactor Containment Building.

Tory II-A: a nuclear ramjet test reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hadley, J.W.

Declassified 28 Nov 1973. The first test reactor in the Pluto program, leading to development of a nuclear ramjet engine, is called Tory II-A. While it is not an actual prototype engine, this reactor embodies a core design which is considered feasible for an engine, and operation of the reactor will provide a test of that core type as well as more generalized values in reactor design and testing. The design of Tory II-A and construction of the reactor and of its test facility are described. Operation of the Tory II-A core at a total power of 160 megawatts, withmore » 800 pounds of air per second passing through the core and emerging at a temperature of 2000 deg F, is the central objective of the test program. All other reactor and facility components exist to support operation of the core, and preliminary steps in the test program itself will be directed primarily toward ensuring attalnment of full-power operation and collection of meaningful data on core behavior during that operation. The core, 3 feet in diameter and 41/2 feet long, will be composed of bundled ceramic tubes whose central holes will provide continuous air passages from end to end of the reactor. These tubes are to be composed of a homogeneous mixture of UO/sub 2/ fuel and BeO moderator, compacted and sintered to achieve high strength and density. (30 references) (auth)« less

REACTOR SERVICE BUILDING, TRA635. CROWDED MOCKUP AREA. CAMERA FACES EAST. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

REACTOR SERVICE BUILDING, TRA-635. CROWDED MOCK-UP AREA. CAMERA FACES EAST. PHOTOGRAPHER'S NOTE SAYS "PICTURE REQUESTED BY IDO IN SUPPORT OF FY '58 BUILDING PROJECTS." INL NEGATIVE NO. 56-3025. R.G. Larsen, Photographer, 9/13/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Proceedings of the AFOSR Special Conference on Prime-Power for High Energy Space Systems, Norfolk, Virginia, 22-25 February 1982. Volume I.

DTIC Science & Technology

1982-02-25

source both liquid and solid fuel combustion devices have been successfully demonstrated during various development programs . Nuclear reactor heat...U02 fuel in the core . Improving the heat pipe model to correlate more closely with the experimental data is a major concern in the development of...ORGANIZATION NAME AND ADDRESS 10. PROGRAM ELEMENT, PROJECT. TASK Research & Development Associates (RDA) AREA &WKNT AE (X Rosslyn, VA 22209 61102F 2301

Proceedings of the first ERDA statistical symposium, Los Alamos, NM, November 3--5, 1975. [Sixteen papers

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nicholson, W L; Harris, J L

1976-03-01

The First ERDA Statistical Symposium was organized to provide a means for communication among ERDA statisticians, and the sixteen papers presented at the meeting are given. Topics include techniques of numerical analysis used for accelerators, nuclear reactors, skewness and kurtosis statistics, radiochemical spectral analysis, quality control, and other statistics problems. Nine of the papers were previously announced in Nuclear Science Abstracts (NSA), while the remaining seven were abstracted for ERDA Energy Research Abstracts (ERA) and INIS Atomindex. (PMA)

Advanced Information Systems Design: Technical Basis and Human Factors Review Guidance

DTIC Science & Technology

2000-03-01

D ., Wise, J ., and Hanes, L., "An Evaluation of Nuclear Power Plant Safety Parameter Display Systems," Proceedings of the Human Factors Society 25th...Reactor (PWR) (Source: Reprinted with permission from Woods, D ., Wise, J ., and Hanes, L., "An Evaluation of Nuclear Power Plant Safety Parameter...Dials display rpCJni?3 (b) Fluid-Tanks display B (c) Seesaw display I 72 CF \\^- J B ’ V ’II ’ ( d ) Mimic display B E * • \\ ^r 7

Proceedings of the conference on energy conservation: Retrofit of Municipal Wastewater Treatment Facilities

NASA Astrophysics Data System (ADS)

1982-02-01

The concepts of the roles of heat and temperature, ventilation and water removal when organic material is assembled into a self insulating mass are summarized. The practical application of these concepts in static pile configuration or the Rutgers process are outlined. The concepts and means of application are relevant to all design approaches whether or not a reactor structure is involved, whether the process is batch or continuous, and whether the mass is agitated or static.

Nuclear systems for space power and propulsion

NASA Technical Reports Server (NTRS)

Klein, M.

1971-01-01

As exploration and utilization of space proceeds through the 1970s, 1980s, and beyond, spacecraft in earth orbit will become increasingly larger, spacecraft will travel deeper into space, and space activities will involve more complex operations. These trends require increasing amounts of energy for power and propulsion. The role to be played by nuclear energy is presented, including plans for deep space missions using radioisotope generators, the reactor power systems for earth orbiting stations and satellites, and the role of nuclear propulsion in space transportation.

Practice and Problems in Language Testing. Proceedings of the International Language Testing Symposium of the Interuniversitare Sprachtestgruppe (4th, Essex, England, September 14-17, 1981).

ERIC Educational Resources Information Center

Culhane, Terry, Ed.; And Others

These proceedings on language testing include the following papers: (1) "The Style Test," by H. Bonheim; (2) "Testing between the Cultures: A Case Study of ESP Testing in Non-idealized Situations," by R. Brunt; (3) "Graded Objectives and Tests in British Schools," by M. Buckby; (4) "The Spanish Attributive Construction SER/ESTAR + Adjective: Real…

ENGINEERING TEST REACTOR, TRA642. CONTEXTUAL VIEW ORIENTATING ETR TO MTR. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ENGINEERING TEST REACTOR, TRA-642. CONTEXTUAL VIEW ORIENTATING ETR TO MTR. CAMERA IS ON ROOF OF MTR BUILDING AND FACES DUE SOUTH. MTR SERVICE BUILDING, TRA-635, IN LOWER RIGHT CORNER. STEEL FRAMES SHOW BUILDINGS TO BE ATTACHED TO ETR BUILDING. HIGH-BAY SECTION IN CENTER IS REACTOR BUILDING. TWO-STORY CONTROL ROOM AND OFFICE BUILDING, TRA-647, IS BETWEEN IT AND MTR SERVICE BUILDING. STRUCTURE TO THE LEFT (WITH NO FRAMING YET) IS COMPRESSOR BUILDING, TRA-643, AND BEYOND IT WILL BE HEAT EXCHANGER BUILDING, TRA-644, GREAT SOUTHERN BUTTE ON HORIZON. INL NEGATIVE NO. 56-2382. Jack L. Anderson, Photographer, 6/10/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

78 FR 64027 - Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-25

...The U.S. Nuclear Regulatory Commission (NRC) is issuing a revision to regulatory guide (RG), 1.79, ``Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors.'' This RG is being revised to incorporate guidance for preoperational testing of new pressurized water reactor (PWR) designs.

78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-24

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0134] Initial Test Program of Emergency Core Cooling....79.1, ``Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors.'' This... emergency core cooling systems (ECCSs) for boiling- water reactors (BWRs) whose licenses are issued after...

Batch Tests To Determine Activity Distribution and Kinetic Parameters for Acetate Utilization in Expanded-Bed Anaerobic Reactors

PubMed Central

Fox, Peter; Suidan, Makram T.

1990-01-01

Batch tests to measure maximum acetate utilization rates were used to determine the distribution of acetate utilizers in expanded-bed sand and expanded-bed granular activated carbon (GAC) reactors. The reactors were fed a mixture of acetate and 3-ethylphenol, and they contained the same predominant aceticlastic methanogen, Methanothrix sp. Batch tests were performed both on the entire reactor contents and with media removed from the reactors. Results indicated that activity was evenly distributed within the GAC reactors, whereas in the sand reactor a sludge blanket on top of the sand bed contained approximately 50% of the activity. The Monod half-velocity constant (Ks) for the acetate-utilizing methanogens in two expanded-bed GAC reactors was searched for by combining steady-state results with batch test data. All parameters necessary to develop a model with Monod kinetics were experimentally determined except for Ks. However, Ks was a function of the effluent 3-ethylphenol concentration, and batch test results demonstrated that maximum acetate utilization rates were not a function of the effluent 3-ethylphenol concentration. Addition of a competitive inhibition term into the Monod expression predicted the dependence of Ks on the effluent 3-ethylphenol concentration. A two-parameter search determined a Ks of 8.99 mg of acetate per liter and a Ki of 2.41 mg of 3-ethylphenol per liter. Model predictions were in agreement with experimental observations for all effluent 3-ethylphenol concentrations. Batch tests measured the activity for a specific substrate and determined the distribution of activity in the reactor. The use of steady-state data in conjunction with batch test results reduced the number of unknown kinetic parameters and thereby reduced the uncertainty in the results and the assumptions made. PMID:16348175

Batch tests to determine activity distribution and kinetic parameters for acetate utilization in expanded-bed anaerobic reactors.

PubMed

Fox, P; Suidan, M T

1990-04-01

Batch tests to measure maximum acetate utilization rates were used to determine the distribution of acetate utilizers in expanded-bed sand and expanded-bed granular activated carbon (GAC) reactors. The reactors were fed a mixture of acetate and 3-ethylphenol, and they contained the same predominant aceticlastic methanogen, Methanothrix sp. Batch tests were performed both on the entire reactor contents and with media removed from the reactors. Results indicated that activity was evenly distributed within the GAC reactors, whereas in the sand reactor a sludge blanket on top of the sand bed contained approximately 50% of the activity. The Monod half-velocity constant (K(s)) for the acetate-utilizing methanogens in two expanded-bed GAC reactors was searched for by combining steady-state results with batch test data. All parameters necessary to develop a model with Monod kinetics were experimentally determined except for K(s). However, K(s) was a function of the effluent 3-ethylphenol concentration, and batch test results demonstrated that maximum acetate utilization rates were not a function of the effluent 3-ethylphenol concentration. Addition of a competitive inhibition term into the Monod expression predicted the dependence of K(s) on the effluent 3-ethylphenol concentration. A two-parameter search determined a K(s) of 8.99 mg of acetate per liter and a K(i) of 2.41 mg of 3-ethylphenol per liter. Model predictions were in agreement with experimental observations for all effluent 3-ethylphenol concentrations. Batch tests measured the activity for a specific substrate and determined the distribution of activity in the reactor. The use of steady-state data in conjunction with batch test results reduced the number of unknown kinetic parameters and thereby reduced the uncertainty in the results and the assumptions made.

Nuclear Reactors. Revised.

ERIC Educational Resources Information Center

Hogerton, John F.

This publication is one of a series of information booklets for the general public published by the United States Atomic Energy Commission. Among the topics discussed are: How Reactors Work; Reactor Design; Research, Teaching, and Materials Testing; Reactors (Research, Teaching and Materials); Production Reactors; Reactors for Electric Power…

49 CFR 40.323 - May program participants release drug or alcohol test information in connection with legal...

Code of Federal Regulations, 2012 CFR

2012-10-01

... collision, the court could determine that a post-accident drug test result of an employee is relevant to... test information in connection with legal proceedings? 40.323 Section 40.323 Transportation Office of... alcohol test information in connection with legal proceedings? (a) As an employer, you may release...

49 CFR 40.323 - May program participants release drug or alcohol test information in connection with legal...

Code of Federal Regulations, 2013 CFR

2013-10-01

... collision, the court could determine that a post-accident drug test result of an employee is relevant to... test information in connection with legal proceedings? 40.323 Section 40.323 Transportation Office of... alcohol test information in connection with legal proceedings? (a) As an employer, you may release...

49 CFR 40.323 - May program participants release drug or alcohol test information in connection with legal...

Code of Federal Regulations, 2011 CFR

2011-10-01

... collision, the court could determine that a post-accident drug test result of an employee is relevant to... test information in connection with legal proceedings? 40.323 Section 40.323 Transportation Office of... alcohol test information in connection with legal proceedings? (a) As an employer, you may release...

49 CFR 40.323 - May program participants release drug or alcohol test information in connection with legal...

Code of Federal Regulations, 2010 CFR

2010-10-01

... collision, the court could determine that a post-accident drug test result of an employee is relevant to... test information in connection with legal proceedings? 40.323 Section 40.323 Transportation Office of... alcohol test information in connection with legal proceedings? (a) As an employer, you may release...

49 CFR 40.323 - May program participants release drug or alcohol test information in connection with legal...

Code of Federal Regulations, 2014 CFR

2014-10-01

... collision, the court could determine that a post-accident drug test result of an employee is relevant to... test information in connection with legal proceedings? 40.323 Section 40.323 Transportation Office of... alcohol test information in connection with legal proceedings? (a) As an employer, you may release...

Thermally Simulated 32kW Direct-Drive Gas-Cooled Reactor: Design, Assembly, and Test

NASA Astrophysics Data System (ADS)

Godfroy, Thomas J.; Kapernick, Richard J.; Bragg-Sitton, Shannon M.

2004-02-01

One of the power systems under consideration for nuclear electric propulsion is a direct-drive gas-cooled reactor coupled to a Brayton cycle. In this system, power is transferred from the reactor to the Brayton system via a circulated closed loop gas. To allow early utilization, system designs must be relatively simple, easy to fabricate, and easy to test using non-nuclear heaters to closely mimic heat from fission. This combination of attributes will allow pre-prototypic systems to be designed, fabricated, and tested quickly and affordably. The ability to build and test units is key to the success of a nuclear program, especially if an early flight is desired. The ability to perform very realistic non-nuclear testing increases the success probability of the system. In addition, the technologies required by a concept will substantially impact the cost, time, and resources required to develop a successful space reactor power system. This paper describes design features, assembly, and test matrix for the testing of a thermally simulated 32kW direct-drive gas-cooled reactor in the Early Flight Fission - Test Facility (EFF-TF) at Marshall Space Flight Center. The reactor design and test matrix are provided by Los Alamos National Laboratories.

GRAPHITE BLOCKS ARE ARRAYED IN "THERMAL COLUMN" ON NORTH SIDE ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

GRAPHITE BLOCKS ARE ARRAYED IN "THERMAL COLUMN" ON NORTH SIDE OF REACTOR. INL NEGATIVE NO. 4000. Unknown Photographer, 12/28/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR BUILDING, TRA642, INTERIOR. FIRST FLOOR. REACTOR IS IN CENTER ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR BUILDING, TRA-642, INTERIOR. FIRST FLOOR. REACTOR IS IN CENTER OF VIEW. CAMERA FACES NORTHWEST. NOTE CRANE RAILS AND DANGLING ELECTRICAL CABLE AT UPPER PART OF VIEW FOR "MOFFETT 2 TON" CRANE. INL NEGATIVE NO. HD46-14-4. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

75 FR 11375 - Revision of Fee Schedules; Fee Recovery for FY 2010

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-10

... Spent Fuel Storage/Reactor Decommissioning..... 2.7 0.2 0.2 Test and Research Reactors 0.2 0.0 0.0 Fuel... categories of licenses. The FY 2009 fee is also shown for comparative purposes. Table V--Rebaselined Annual...) Spent Fuel Storage/Reactor 122,000 143,000 Decommissioning Test and Research Reactors (Non-power 87,600...

ETR, TRA642. ON BASEMENT FLOOR. REACTOR VESSEL WILL BE PLACED ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR, TRA-642. ON BASEMENT FLOOR. REACTOR VESSEL WILL BE PLACED WITHIN THE INNER METAL FORM. WHEN CONCRETE IS POURED OUTSIDE THIS FORM, CONDUIT HOLES WILL BE PRESERVE SPACE THROUGH HOLES. INL NEGATIVE NO. 56-1507. Jack L. Anderson, Photographer, 5/8/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

REACTIVITY MEASUREMENT FACILITY. CAMERA LOOKS DOWN INTO MTR CANAL. REACTOR ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

REACTIVITY MEASUREMENT FACILITY. CAMERA LOOKS DOWN INTO MTR CANAL. REACTOR IS FUELED AS AN ETR MOCK-UP. LIGHTS DANGLE BELOW WATER LEVEL. CONTROL RODS AND OTHER APPARATUS DESCEND FROM ABOVE WATER LEVEL. INL NEGATIVE NO. 56-900. Jack L. Anderson, Photographer, 3/26/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

REACTOR SERVICES BUILDING, TRA635, INTERIOR. ALSO KNOWN AS MATERIAL RECEIVING ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

REACTOR SERVICES BUILDING, TRA-635, INTERIOR. ALSO KNOWN AS MATERIAL RECEIVING AREA AND LABORATORY. CAMERA ON FIRST FLOOR FACING NORTH TOWARD MTR BUILDING. MOCK-UP AREA WAS TO THE RIGHT OF VIEW. INL NEGATIVE NO. HD46-10-1. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Safety Issues at the DOE Test and Research Reactors. A Report to the U.S. Department of Energy.

ERIC Educational Resources Information Center

National Academy of Sciences - National Research Council, Washington, DC. Commission on Physical Sciences, Mathematics, and Resources.

This report provides an assessment of safety issues at the Department of Energy (DOE) test and research reactors. Part A identifies six safety issues of the reactors. These issues include the safety design philosophy, the conduct of safety reviews, the performance of probabilistic risk assessments, the reliance on reactor operators, the fragmented…

Implicit time-integration method for simultaneous solution of a coupled non-linear system

NASA Astrophysics Data System (ADS)

Watson, Justin Kyle

Historically large physical problems have been divided into smaller problems based on the physics involved. This is no different in reactor safety analysis. The problem of analyzing a nuclear reactor for design basis accidents is performed by a handful of computer codes each solving a portion of the problem. The reactor thermal hydraulic response to an event is determined using a system code like TRAC RELAP Advanced Computational Engine (TRACE). The core power response to the same accident scenario is determined using a core physics code like Purdue Advanced Core Simulator (PARCS). Containment response to the reactor depressurization in a Loss Of Coolant Accident (LOCA) type event is calculated by a separate code. Sub-channel analysis is performed with yet another computer code. This is just a sample of the computer codes used to solve the overall problems of nuclear reactor design basis accidents. Traditionally each of these codes operates independently from each other using only the global results from one calculation as boundary conditions to another. Industry's drive to uprate power for reactors has motivated analysts to move from a conservative approach to design basis accident towards a best estimate method. To achieve a best estimate calculation efforts have been aimed at coupling the individual physics models to improve the accuracy of the analysis and reduce margins. The current coupling techniques are sequential in nature. During a calculation time-step data is passed between the two codes. The individual codes solve their portion of the calculation and converge to a solution before the calculation is allowed to proceed to the next time-step. This thesis presents a fully implicit method of simultaneous solving the neutron balance equations, heat conduction equations and the constitutive fluid dynamics equations. It discusses the problems involved in coupling different physics phenomena within multi-physics codes and presents a solution to these problems. The thesis also outlines the basic concepts behind the nodal balance equations, heat transfer equations and the thermal hydraulic equations, which will be coupled to form a fully implicit nonlinear system of equations. The coupling of separate physics models to solve a larger problem and improve accuracy and efficiency of a calculation is not a new idea, however implementing them in an implicit manner and solving the system simultaneously is. Also the application to reactor safety codes is new and has not be done with thermal hydraulics and neutronics codes on realistic applications in the past. The coupling technique described in this thesis is applicable to other similar coupled thermal hydraulic and core physics reactor safety codes. This technique is demonstrated using coupled input decks to show that the system is solved correctly and then verified by using two derivative test problems based on international benchmark problems the OECD/NRC Three mile Island (TMI) Main Steam Line Break (MSLB) problem (representative of pressurized water reactor analysis) and the OECD/NRC Peach Bottom (PB) Turbine Trip (TT) benchmark (representative of boiling water reactor analysis).

The results of systems tests of the 500 kV busbar controllable shunting reactor in the Tavricheskaya substation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gusev, S. I.; Karpov, V. N.; Kiselev, A. N.

2009-09-15

The results of systems tests of the 500 kV busbar magnetization-controllable shunting reactor (CSR), set up in the Tavricheskaya substation, including measurements of the quality of the electric power, the harmonic composition of the network currents of the reactor for different values of the reactive power consumed, the determination of the regulating characteristics of the reactor, the speed of response of the shunting reactor in the current and voltage stabilization modes, and also the operation of the reactor under dynamic conditions for different perturbations, are presented. The results obtained are analyzed.

The Use of Case History Studies to Differentiate Potentially Infected from Potentially Noninfected Herds with Reactors to Brucella abortus Antigens

PubMed Central

Martin, S. W.; Gerrow, A. F.

1978-01-01

Data on farm characteristics and the results of the first herd test for brucellosis were collected for 74 reactor and 74 negative herds in Wellington County, Ontario. Each reactor herd was classified as either probably infected or probably not infected using the occurrence of abortions prior to the first herd test or during the testing period, the total number of cattle removed and/or the spread of reactors within the herd as criteria of infection. Statistical techniques were used to select variables which were good “discriminators” between probably infected and noninfected herds. In general, reactor herds were primarily dairy herds and were somewhat larger than negative herds. The presence of only single suspicious reactors on the first test appeared to be a good predictor of lack of infection with Brucella abortus. Among the 37 farms in this category the single reactor was removed from only eight farms and no evidence o fthe spread of infection was observed. The presence of one or more positive reactors on the first herd test appeared to be a good predictor of the presence of infection. Of the 24 farms in this category, evidence of the spread of infection was present in ten farms and seven of these ten farms were eventually depopulated. The brucella milk ring test appeared to be the most effective means of identifying infected herds under the conditions present in Wellington County. PMID:417777

10 CFR 52.167 - Issuance of manufacturing license.

Code of Federal Regulations, 2010 CFR

2010-01-01

... proposed reactor(s) can be incorporated into a nuclear power plant and operated at sites having... design and manufacture the proposed nuclear power reactor(s); (5) The proposed inspections, tests... the construction of a nuclear power facility using the manufactured reactor(s). (2) A holder of a...

DEVICE FOR TREATING MATERIALS

DOEpatents

Ohlinger, L.A.; Seitz, F.; Young, G.J.

1959-02-17

Test-hole construction in a reactor to facilitate inserting and removing test specimens from the reactor for irradiation therein is discussed. An elongated chamber extends from the outer face of the reactor shield into the reactor. A shield box, having an open end, is sealed to thc outer face of the reactor shield by its open end surrounding the outer end of the chamber. A removable door is provided in the side wall of the shield box for inscrtion and removal of test specimens. A means operable from thc exterior of the shield box is provided for transferring test specimens between the shield box and the irradiation position within the chamber and consists of an elongated rod having a specimen tray engaging member on its inner end, which may be manipulated by the operator.

Report on FY15 alloy 617 code rules development

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sham, Sam; Jetter, Robert I; Hollinger, Greg

2015-09-01

Due to its strength at very high temperatures, up to 950°C (1742°F), Alloy 617 is the reference construction material for structural components that operate at or near the outlet temperature of the very high temperature gas-cooled reactors. However, the current rules in the ASME Section III, Division 5 Subsection HB, Subpart B for the evaluation of strain limits and creep-fatigue damage using simplified methods based on elastic analysis have been deemed inappropriate for Alloy 617 at temperatures above 650°C (1200°F) (Corum and Brass, Proceedings of ASME 1991 Pressure Vessels and Piping Conference, PVP-Vol. 215, p.147, ASME, NY, 1991). The rationalemore » for this exclusion is that at higher temperatures it is not feasible to decouple plasticity and creep, which is the basis for the current simplified rules. This temperature, 650°C (1200°F), is well below the temperature range of interest for this material for the high temperature gas-cooled reactors and the very high temperature gas-cooled reactors. The only current alternative is, thus, a full inelastic analysis requiring sophisticated material models that have not yet been formulated and verified. To address these issues, proposed code rules have been developed which are based on the use of elastic-perfectly plastic (EPP) analysis methods applicable to very high temperatures. The proposed rules for strain limits and creep-fatigue evaluation were initially documented in the technical literature (Carter, Jetter and Sham, Proceedings of ASME 2012 Pressure Vessels and Piping Conference, papers PVP 2012 28082 and PVP 2012 28083, ASME, NY, 2012), and have been recently revised to incorporate comments and simplify their application. Background documents have been developed for these two code cases to support the ASME Code committee approval process. These background documents for the EPP strain limits and creep-fatigue code cases are documented in this report.« less

In-reactor performance of LWR-type tritium target rods

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lanning, D.D.; Paxton, M.M.; Crumbaugh, L.

Pacific Northwest Laboratory has conducted several 1-yr irradiation tests of light water reactor-type tritium target rods. These tests have been sponsored by the U.S. Department of Energy's Office of New Production Reactors. The first test, designated water capsule-1 (WC-1), was conducted in the Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory from November 1989 to December 1990. The test vehicle contained a single 4-ft target rod within a pressurized water capsule. The capsule maintained the rod at pressurized water reactor (PWR)-type water temperature and pressure conditions. Posttest nondestructive examinations of the WC-1 rod involved visual examinations, dimensional checks,more » gamma scanning, and neutron radiography. The results indicate that the rod maintained the integrity of its pressure seal and was otherwise unaltered both mechanically and dimensionally by its irradiation and posttest handling.« less

Observed Changes in As-Fabricated U-10Mo Monolithic Fuel Microstructures After Irradiation in the Advanced Test Reactor

NASA Astrophysics Data System (ADS)

Keiser, Dennis; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Madden, James

2017-12-01

A low-enriched uranium U-10Mo monolithic nuclear fuel is being developed by the Material Management and Minimization Program, earlier known as the Reduced Enrichment for Research and Test Reactors Program, for utilization in research and test reactors around the world that currently use high-enriched uranium fuels. As part of this program, reactor experiments are being performed in the Advanced Test Reactor. It must be demonstrated that this fuel type exhibits mechanical integrity, geometric stability, and predictable behavior to high powers and high fission densities in order for it to be a viable fuel for qualification. This paper provides an overview of the microstructures observed at different regions of interest in fuel plates before and after irradiation for fuel samples that have been tested. These fuel plates were fabricated using laboratory-scale fabrication methods. Observations regarding how microstructural changes during irradiation may impact fuel performance are discussed.

Application of Simulated Reactivity Feedback in Nonnuclear Testing of a Direct-Drive Gas-Cooled Reactor

NASA Technical Reports Server (NTRS)

Bragg-Sitton, S. M.; Webster, K. L.

2007-01-01

Nonnuclear testing can be a valuable tool in the development of an in-space nuclear power or propulsion system. In a nonnuclear test facility, electric heaters are used to simulate heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response and response characteristics, and assess potential design improvements with a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE 100a heat pipe cooled, electrically heated reactor and heat exchanger hardware. This Technical Memorandum discusses the status of the planned dynamic test methodology for implementation in the direct-drive gas-cooled reactor testing and assesses the additional instrumentation needed to implement high-fidelity dynamic testing.

10 CFR 55.5 - Communications.

Code of Federal Regulations, 2012 CFR

2012-01-01

.... Nuclear Regulatory Commission, Washington, DC 20555-0001. The guidance discusses, among other topics, the.... (b)(1) Except for test and research reactor facilities, the Director, Office of Nuclear Reactor... involving a test and research reactor facility licensed under 10 CFR part 50 and any related inquiry...

10 CFR 55.5 - Communications.

Code of Federal Regulations, 2013 CFR

2013-01-01

.... Nuclear Regulatory Commission, Washington, DC 20555-0001. The guidance discusses, among other topics, the.... (b)(1) Except for test and research reactor facilities, the Director, Office of Nuclear Reactor... involving a test and research reactor facility licensed under 10 CFR part 50 and any related inquiry...

Validation of large-scale, monochromatic UV disinfection systems for drinking water using dyed microspheres.

PubMed

Blatchley, E R; Shen, C; Scheible, O K; Robinson, J P; Ragheb, K; Bergstrom, D E; Rokjer, D

2008-02-01

Dyed microspheres have been developed as a new method for validation of ultraviolet (UV) reactor systems. When properly applied, dyed microspheres allow measurement of the UV dose distribution delivered by a photochemical reactor for a given operating condition. Prior to this research, dyed microspheres had only been applied to a bench-scale UV reactor. The goal of this research was to extend the application of dyed microspheres to large-scale reactors. Dyed microsphere tests were conducted on two prototype large-scale UV reactors at the UV Validation and Research Center of New York (UV Center) in Johnstown, NY. All microsphere tests were conducted under conditions that had been used previously in biodosimetry experiments involving two challenge bacteriophage: MS2 and Qbeta. Numerical simulations based on computational fluid dynamics and irradiance field modeling were also performed for the same set of operating conditions used in the microspheres assays. Microsphere tests on the first reactor illustrated difficulties in sample collection and discrimination of microspheres against ambient particles. Changes in sample collection and work-up were implemented in tests conducted on the second reactor that allowed for improvements in microsphere capture and discrimination against the background. Under these conditions, estimates of the UV dose distribution from the microspheres assay were consistent with numerical simulations and the results of biodosimetry, using both challenge organisms. The combined application of dyed microspheres, biodosimetry, and numerical simulation offers the potential to provide a more in-depth description of reactor performance than any of these methods individually, or in combination. This approach also has the potential to substantially reduce uncertainties in reactor validation, thereby leading to better understanding of reactor performance, improvements in reactor design, and decreases in reactor capital and operating costs.

Minutes of the third annual meeting of the Panel on Reference Nuclear Data. [BNL, October 5, 1978

DOE Office of Scientific and Technical Information (OSTI.GOV)

Burrows, T.W.; Stewart, L.; Coyne, J.J.

1979-05-01

The major activities of the meeting were as follows: welcome; organization, approval of minutes of the second meeting, and approval of agenda; review of nuclear data compilation and evaluation efforts (national and international efforts, master data files, publications); summary of 1977 panel meeting; definition of reference nuclear data; discussion of specific data needs and possible data center contributions (reactor physics, medicine and biology, controlled thermonuclear reactors and astrophysics); establishment of current interest and future direction of the panel; adjournment. Recommendations and action items are listed. Tables on nuclear data needs in applied physics, medicine and biology, and controlled thermonuclear reactorsmore » and astrophysics are presented. Appendixes include membership lists of various committees, summaries of publication activities, survey results, correspondence, and portions of the documents Proceedings of the Magnetic Fusion Energy Blanket and Shield Workshop and National Needs for Critically Evaluated Physical and Chemical Data. (RWR)« less

Status of the irradiation test vehicle for testing fusion materials in the Advanced Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tsai, H.; Gomes, I.C.; Smith, D.L.

1998-09-01

The design of the irradiation test vehicle (ITV) for the Advanced Test Reactor (ATR) has been completed. The main application for the ITV is irradiation testing of candidate fusion structural materials, including vanadium-base alloys, silicon carbide composites, and low-activation steels. Construction of the vehicle is underway at the Lockheed Martin Idaho Technology Company (LMITCO). Dummy test trains are being built for system checkout and fine-tuning. Reactor insertion of the ITV with the dummy test trains is scheduled for fall 1998. Barring unexpected difficulties, the ITV will be available for experiments in early 1999.

10 CFR 55.5 - Communications.

Code of Federal Regulations, 2011 CFR

2011-01-01

.... Nuclear Regulatory Commission, Washington, DC 20555-0001. The guidance discusses, among other topics, the.... (b)(1) Except for test and research reactor facilities, the Director, Office of Nuclear Reactor... this part involving a test and research reactor facility licensed under 10 CFR part 50 and any related...

TREAT Reactor Control and Protection System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lipinski, W.C.; Brookshier, W.K.; Burrows, D.R.

1985-01-01

The main control algorithm of the Transient Reactor Test Facility (TREAT) Automatic Reactor Control System (ARCS) resides in Read Only Memory (ROM) and only experiment specific parameters are input via keyboard entry. Prior to executing an experiment, the software and hardware of the control computer is tested by a closed loop real-time simulation. Two computers with parallel processing are used for the reactor simulation and another computer is used for simulation of the control rod system. A monitor computer, used as a redundant diverse reactor protection channel, uses more conservative setpoints and reduces challenges to the Reactor Trip System (RTS).more » The RTS consists of triplicated hardwired channels with one out of three logic. The RTS is automatically tested by a digital Dedicated Microprocessor Tester (DMT) prior to the execution of an experiment. 6 refs., 5 figs., 1 tab.« less

Characteristics of potential repository wastes: Volume 4, Appendix 4A, Nuclear reactors at educational institutions of the United States; Appendix 4B, Data sheets for nuclear reactors at educational institutions; Appendix 4C, Supplemental data for Fort St. Vrain spent fuel; Appendix 4D, Supplemental data for Peach Bottom 1 spent fuel; Appendix 4E, Supplemental data for Fast Flux Test Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1992-07-01

Volume 4 contains the following appendices: nuclear reactors at educational institutions in the United States; data sheets for nuclear reactors at educational institutions in the United States(operational reactors and shut-down reactors); supplemental data for Fort St. Vrain spent fuel; supplemental data for Peach Bottom 1 spent fuel; and supplemental data for Fast Flux Test Facility.

ADAPTATION OF CRACK GROWTH DETECTION TECHNIQUES TO US MATERIAL TEST REACTORS

DOE Office of Scientific and Technical Information (OSTI.GOV)

A. Joseph Palmer; Sebastien P. Teysseyre; Kurt L. Davis

2015-04-01

A key component in evaluating the ability of Light Water Reactors to operate beyond 60 years is characterizing the degradation of materials exposed to radiation and various water chemistries. Of particular concern is the response of reactor materials to Irradiation Assisted Stress Corrosion Cracking (IASCC). Some test reactors outside the United States, such as the Halden Boiling Water Reactor (HBWR), have developed techniques to measure crack growth propagation during irradiation. The basic approach is to use a custom-designed compact loading mechanism to stress the specimen during irradiation, while the crack in the specimen is monitored in-situ using the Direct Currentmore » Potential Drop (DCPD) method. In 2012 the US Department of Energy commissioned the Idaho National Laboratory and the MIT Nuclear Reactor Laboratory (MIT NRL) to take the basic concepts developed at the HBWR and adapt them to a test rig capable of conducting in-pile IASCC tests in US Material Test Reactors. The first two and half years of the project consisted of designing and testing the loader mechanism, testing individual components of the in-pile rig and electronic support equipment, and autoclave testing of the rig design prior to insertion in the MIT Reactor. The load was applied to the specimen by means of a scissor like mechanism, actuated by a miniature metal bellows driven by pneumatic pressure and sized to fit within the small in-core irradiation volume. In addition to the loader design, technical challenges included developing robust connections to the specimen for the applied current and voltage measurements, appropriate ceramic insulating materials that can endure the LWR environment, dealing with the high electromagnetic noise environment of a reactor core at full power, and accommodating material property changes in the specimen, due primarily to fast neutron damage, which change the specimen resistance without additional crack growth. The project culminated with an in-pile demonstration at the MIT Reactor. The test rig and associated support equipment were used to apply loads to a representative Compact Tensile specimen during one MITR operating cycle, while measuring crack growth using the DCPD method. Although the test period was short (approximately 70 days), and the accumulated neutron dose relatively small, successful operation of the test rig was demonstrated. The specimen was cycled more than 8000 times (more than would be typical for a long term IASCC test), which was sufficient to propagate a crack of over 2 mm.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

Digital instrumentation and controls system technique is being introduced in new constructed research reactor or life extension of older research reactor. Digital systems are easy to change and optimize but the validated process for them is required. Also, to reduce project risk or cost, we have to make it sure that configuration and control functions is right before the commissioning phase on research reactor. For this purpose, simulators have been widely used in developing control systems in automotive and aerospace industries. In these literatures, however, very few of these can be found regarding test on the control system of researchmore » reactor with simulator. Therefore, this paper proposes a simulation platform to verify the performance of RRS (Reactor Regulating System) for research reactor. This simulation platform consists of the reactor simulation model and the interface module. This simulation platform is applied to I and C upgrade project of TRIGA reactor, and many problems of RRS configuration were found and solved. And it proved that the dynamic performance testing based on simulator enables significant time saving and improves economics and quality for RRS in the system test phase. (authors)« less

9 CFR 51.4 - Record of tests.

Code of Federal Regulations, 2013 CFR

2013-01-01

... COOPERATIVE CONTROL AND ERADICATION OF LIVESTOCK OR POULTRY DISEASES ANIMALS DESTROYED BECAUSE OF BRUCELLOSIS... affected, including the reactor tag number of each brucellosis reactor animal and the registration name and number of each brucellosis reactor registered animal. A copy of the applicable test record shall be given...

9 CFR 51.4 - Record of tests.

Code of Federal Regulations, 2011 CFR

2011-01-01

... COOPERATIVE CONTROL AND ERADICATION OF LIVESTOCK OR POULTRY DISEASES ANIMALS DESTROYED BECAUSE OF BRUCELLOSIS... affected, including the reactor tag number of each brucellosis reactor animal and the registration name and number of each brucellosis reactor registered animal. A copy of the applicable test record shall be given...

9 CFR 51.4 - Record of tests.

Code of Federal Regulations, 2012 CFR

2012-01-01

... COOPERATIVE CONTROL AND ERADICATION OF LIVESTOCK OR POULTRY DISEASES ANIMALS DESTROYED BECAUSE OF BRUCELLOSIS... affected, including the reactor tag number of each brucellosis reactor animal and the registration name and number of each brucellosis reactor registered animal. A copy of the applicable test record shall be given...

9 CFR 51.4 - Record of tests.

Code of Federal Regulations, 2014 CFR

2014-01-01

... COOPERATIVE CONTROL AND ERADICATION OF LIVESTOCK OR POULTRY DISEASES ANIMALS DESTROYED BECAUSE OF BRUCELLOSIS... affected, including the reactor tag number of each brucellosis reactor animal and the registration name and number of each brucellosis reactor registered animal. A copy of the applicable test record shall be given...

Demonstration of Robustness and Integrated Operation of a Series-Bosch System

NASA Technical Reports Server (NTRS)

Abney, Morgan B.; Mansell, Matthew J.; Stanley, Christine; Barnett, Bill; Junaedi, Christian; Vilekar, Saurabh A.; Ryan, Kent

2016-01-01

Manned missions beyond low Earth orbit will require highly robust, reliable, and maintainable life support systems that maximize recycling of water and oxygen. Bosch technology is one option to maximize oxygen recovery, in the form of water, from metabolically-produced carbon dioxide (CO2). A two stage approach to Bosch, called Series-Bosch, reduces metabolic CO2 with hydrogen (H2) to produce water and solid carbon using two reactors: a Reverse Water-Gas Shift (RWGS) reactor and a carbon formation (CF) reactor. Previous development efforts demonstrated the stand-alone performance of a NASA-designed RWGS reactor designed for robustness against carbon formation, two membrane separators intended to maximize single pass conversion of reactants, and a batch CF reactor with both transit and surface catalysts. In the past year, Precision Combustion, Inc. (PCI) developed and delivered a RWGS reactor for testing at NASA. The reactor design was based on their patented Microlith® technology and was first evaluated under a Phase I Small Business Innovative Research (SBIR) effort in 2010. The RWGS reactor was recently evaluated at NASA to compare its performance and operating conditions with NASA's RWGS reactor. The test results will be provided in this paper. Separately, in 2015, a semi-continuous CF reactor was designed and fabricated at NASA based on the results from batch CF reactor testing. The batch CF reactor and the semi-continuous CF reactor were individually integrated with an upstream RWGS reactor to demonstrate the system operation and to evaluate performance. Here, we compare the performance and robustness to carbon formation of both RWGS reactors. We report the results of the integrated operation of a Series-Bosch system and we discuss the technology readiness level.

A liquid-metal filling system for pumped primary loop space reactors

NASA Astrophysics Data System (ADS)

Crandall, D. L.; Reed, W. C.

Some concepts for the SP-100 space nuclear power reactor use liquid metal as the primary coolant in a pumped loop. Prior to filling ground engineering test articles or reactor systems, the liquid metal must be purified and circulated through the reactor primary system to remove contaminants. If not removed, these contaminants enhance corrosion and reduce reliability. A facility was designed and built to support Department of Energy Liquid Metal Fast Breeder Reactor tests conducted at the Idaho National Engineering Laboratory. This test program used liquid sodium to cool nuclear fuel in in-pile experiments; thus, a system was needed to store and purify sodium inventories and fill the experiment assemblies. This same system, with modifications and potential changeover to lithium or sodium-potassium (NaK), can be used in the Space Nuclear Power Reactor Program. This paper addresses the requirements, description, modifications, operation, and appropriateness of using this liquid-metal system to support the SP-100 space reactor program.

10 CFR 171.3 - Scope.

Code of Federal Regulations, 2010 CFR

2010-01-01

... holding an operating license for a power reactor, test reactor or research reactor issued under part 50 of... authorizes operation of a power reactor. The regulations in this part also apply to any person holding a...

Advanced Instrumentation for Transient Reactor Testing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Corradini, Michael L.; Anderson, Mark; Imel, George

Transient testing involves placing fuel or material into the core of specialized materials test reactors that are capable of simulating a range of design basis accidents, including reactivity insertion accidents, that require the reactor produce short bursts of intense highpower neutron flux and gamma radiation. Testing fuel behavior in a prototypic neutron environment under high-power, accident-simulation conditions is a key step in licensing nuclear fuels for use in existing and future nuclear power plants. Transient testing of nuclear fuels is needed to develop and prove the safety basis for advanced reactors and fuels. In addition, modern fuel development and designmore » increasingly relies on modeling and simulation efforts that must be informed and validated using specially designed material performance separate effects studies. These studies will require experimental facilities that are able to support variable scale, highly instrumented tests providing data that have appropriate spatial and temporal resolution. Finally, there are efforts now underway to develop advanced light water reactor (LWR) fuels with enhanced performance and accident tolerance. These advanced reactor designs will also require new fuel types. These new fuels need to be tested in a controlled environment in order to learn how they respond to accident conditions. For these applications, transient reactor testing is needed to help design fuels with improved performance. In order to maximize the value of transient testing, there is a need for in-situ transient realtime imaging technology (e.g., the neutron detection and imaging system like the hodoscope) to see fuel motion during rapid transient excursions with a higher degree of spatial and temporal resolution and accuracy. There also exists a need for new small, compact local sensors and instrumentation that are capable of collecting data during transients (e.g., local displacements, temperatures, thermal conductivity, neutron flux, etc.).« less

Integrated System for Retrieval, Transportation and Consolidated Storage of Used Nuclear Fuel in the US - 13312

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bracey, William; Bondre, Jayant; Shelton, Catherine

2013-07-01

The current inventory of used nuclear fuel assemblies (UNFAs) from commercial reactor operations in the United States totals approximately 65,000 metric tons or approximately 232,000 UNFAs primarily stored at the 104 operational reactors in the US and a small number of decommissioned reactors. This inventory is growing at a rate of roughly 2,000 to 2,400 metric tons each year, (Approx. 7,000 UNFAs) as a result of ongoing commercial reactor operations. Assuming an average of 10 metric tons per storage/transportation casks, this inventory of commercial UNFAs represents about 6,500 casks with an additional of about 220 casks every year. In Januarymore » 2010, the Blue Ribbon Commission (BRC) [1] was directed to conduct a comprehensive review of policies for managing the back end of the nuclear fuel cycle and recommend a new plan. The BRC issued their final recommendations in January 2012. One of the main recommendations is for the United States to proceed promptly to develop one or more consolidated storage facilities (CSF) as part of an integrated, comprehensive plan for safely managing the back end of the nuclear fuel cycle. Based on its extensive experience in storage and transportation cask design, analysis, licensing, fabrication, and operations including transportation logistics, Transnuclear, Inc. (TN), an AREVA Subsidiary within the Logistics Business Unit, is engineering an integrated system that will address the complete process of commercial UNFA management. The system will deal with UNFAs in their current storage mode in various configurations, the preparation including handling and additional packaging where required and transportation of UNFAs to a CSF site, and subsequent storage, operation and maintenance at the CSF with eventual transportation to a future repository or recycling site. It is essential to proceed by steps to ensure that the system will be the most efficient and serve at best its purpose by defining: the problem to be resolved, the criteria to evaluate the solutions, and the alternative solutions. The complexity of the project is increasing with time (more fuel assemblies, new storage systems, deteriorating logistics infrastructure at some sites, etc.) but with the uncertainty on the final disposal path, flexibility and simplicity will be critical. (authors)« less

REACTOR SERVICE BUILDING, TRA635, CONTEXTUAL VIEW DURING CONSTRUCTION. CAMERA IS ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

REACTOR SERVICE BUILDING, TRA-635, CONTEXTUAL VIEW DURING CONSTRUCTION. CAMERA IS ATOP MTR BUILDING AND LOOKING SOUTHERLY. FOUNDATION AND DRAINS ARE UNDER CONSTRUCTION. THE BUILDING WILL BUTT AGAINST CHARGING FACE OF PLUG STORAGE BUILDING. HOT CELL BUILDING, TRA-632, IS UNDER CONSTRUCTION AT TOP CENTER OF VIEW. INL NEGATIVE NO. 8518. Unknown Photographer, 8/25/1953 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Reactor transient control in support of PFR/TREAT TUCOP experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Burrows, D.R.; Larsen, G.R.; Harrison, L.J.

1984-01-01

Unique energy deposition and experiment control requirements posed bythe PFR/TREAT series of transient undercooling/overpower (TUCOP) experiments resulted in equally unique TREAT reactor operations. New reactor control computer algorithms were written and used with the TREAT reactor control computer system to perform such functions as early power burst generation (based on test train flow conditions), burst generation produced by a step insertion of reactivity following a controlled power ramp, and shutdown (SCRAM) initiators based on both test train conditions and energy deposition. Specialized hardware was constructed to simulate test train inputs to the control computer system so that computer algorithms couldmore » be tested in real time without irradiating the experiment.« less

49 CFR 40.253 - What are the procedures for conducting an alcohol confirmation test?

Code of Federal Regulations, 2010 CFR

2010-10-01

... confirmation test and show the reading to the employee. (1) If the reading is 0.00, the test may proceed. If the reading is greater than 0.00, you must conduct another air blank. (2) If the reading on the second air blank is 0.00, the test may proceed. If the reading is greater than 0.00, you must take the EBT...

49 CFR 40.253 - What are the procedures for conducting an alcohol confirmation test?

Code of Federal Regulations, 2011 CFR

2011-10-01

... confirmation test and show the reading to the employee. (1) If the reading is 0.00, the test may proceed. If the reading is greater than 0.00, you must conduct another air blank. (2) If the reading on the second air blank is 0.00, the test may proceed. If the reading is greater than 0.00, you must take the EBT...

Hardening neutron spectrum for advanced actinide transmutation experiments in the ATR.

PubMed

Chang, G S; Ambrosek, R G

2005-01-01

The most effective method for transmuting long-lived isotopes contained in spent nuclear fuel into shorter-lived fission products is in a fast neutron spectrum reactor. In the absence of a fast test reactor in the United States, initial irradiation testing of candidate fuels can be performed in a thermal test reactor that has been modified to produce a test region with a hardened neutron spectrum. Such a test facility, with a spectrum similar but somewhat softer than that of the liquid-metal fast breeder reactor (LMFBR), has been constructed in the INEEL's Advanced Test Reactor (ATR). The radial fission power distribution of the actinide fuel pin, which is an important parameter in fission gas release modelling, needs to be accurately predicted and the hardened neutron spectrum in the ATR and the LMFBR fast neutron spectrum is compared. The comparison analyses in this study are performed using MCWO, a well-developed tool that couples the Monte Carlo transport code MCNP with the isotope depletion and build-up code ORIGEN-2. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations and detailed radial fission power profile calculations for a typical fast reactor (LMFBR) neutron spectrum and the hardened neutron spectrum test region in the ATR. The MCWO-calculated results indicate that the cadmium basket used in the advanced fuel test assembly in the ATR can effectively depress the linear heat generation rate in the experimental fuels and harden the neutron spectrum in the test region.

Proceedings of the International Congress (12th), Corrosion Control for Low-Cost Reliability, Held in Houston, Texas on September 19 -24, 1993. Volume 1. Coatings

DTIC Science & Technology

1993-09-24

3]) Gas-cooled reactors were first developed in Europe and have been built since 1956. HTGR , equipped with the core of ceramic coated particle fuels ...demands must also be covered by nuclear energy in not so long future. Programs on developing the process heating HTGR have been promoted mainly in Germany...Material programs for HTGR have been promoted in several countries since late 1960’s which include the tasks of developing and qualifying materials, eg

Optics in engineering measurement; Proceedings of the Meeting, Cannes, France, December 3-6, 1985

NASA Technical Reports Server (NTRS)

Fagan, William F. (Editor)

1986-01-01

The present conference on optical measurement systems considers topics in the fields of holographic interferometry, speckle techniques, moire fringe and grating methods, optical surface gaging, laser- and fiber-optics-based measurement systems, and optics for engineering data evaluation. Specific attention is given to holographic NDE for aerospace composites, holographic interferometry of rotating components, new developments in computer-aided holography, electronic speckle pattern interferometry, mass transfer measurements using projected fringes, nuclear reactor photogrammetric inspection, a laser Doppler vibrometer, and optoelectronic measurements of the yaw angle of projectiles.

Nuclear Science Symposium, 4th, and Nuclear Power Systems Symposium, 9th, San Francisco, Calif., October 19-21, 1977, Proceedings

NASA Technical Reports Server (NTRS)

1978-01-01

Consideration is given to the following types of high energy physics instrumentation: drift chambers, multiwire proportional chambers, calorimeters, optical detectors, ionization and scintillation detectors, solid state detectors, and electronic and digital subsystems. Attention is also paid to reactor instrumentation, nuclear medicine instrumentation, data acquisition systems for nuclear instrumentation, microprocessor applications in nuclear science, environmental instrumentation, control and instrumentation of nuclear power generating stations, and radiation monitoring. Papers are also presented on instrumentation for the High Energy Astronomy Observatory.

Long-Baseline Neutrino Experiments

DOE PAGES

Diwan, M. V.; Galymov, V.; Qian, X.; ...

2016-10-19

We review long-baseline neutrino experiments in which neutrinos are detected after traversing macroscopic distances. Over such distances neutrinos have been found to oscillate among flavor states. Experiments with solar, atmospheric, reactor, and accelerator neutrinos have resulted in a coherent picture of neutrino masses and mixing of the three known flavor states. We will summarize the current best knowledge of neutrino parameters and phenomenology with our focus on the evolution of the experimental technique. We will proceed from the rst evidence produced by astrophysical neutrino sources to the current open questions and the goals of future research

HOT CELL BUILDING, TRA632. CONTEXTUAL VIEW ALONG WALLEYE AVENUE, CAMERA ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

HOT CELL BUILDING, TRA-632. CONTEXTUAL VIEW ALONG WALLEYE AVENUE, CAMERA FACING EASTERLY. HOT CELL BUILDING IS AT CENTER LEFT OF VIEW; THE LOW-BAY PROJECTION WITH LADDER IS THE TEST TRAIN ASSEMBLY FACILITY, ADDED IN 1968. MTR BUILDING IS IN LEFT OF VIEW. HIGH-BAY BUILDING AT RIGHT IS THE ENGINEERING TEST REACTOR BUILDING, TRA-642. INL NEGATIVE NO. HD46-32-1. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Proceedings: Nozzle Initiative Industry Advisory Committee on Standardization of Carbon-Phenolic Test Methods and Specifications

NASA Technical Reports Server (NTRS)

1992-01-01

The proceedings of the meeting is presented in conversational form. Some areas of discussion are as follow: resin advancement at NASA Marshall new technologies studies; NMR studies; SPIP/PAN development summary; computer modeling support; composite testing; carbon assay testing; activity and aerospace computer database; alternate rayon yarn sizing; fiber morphology; and carbon microballoons specifications.

Proposed Design and Operation of a Heat Pipe Reactor using the Sandia National Laboratories Annular Core Test Facility and Existing UZrH Fuel Pins

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wright, Steven A.; Lipinski, Ronald J.; Pandya, Tara

2005-02-06

Heat Pipe Reactors (HPR) for space power conversion systems offer a number of advantages not easily provided by other systems. They require no pumping, their design easily deals with freezing and thawing of the liquid metal, and they can provide substantial levels of redundancy. Nevertheless, no reactor has ever been operated and cooled with heat pipes, and the startup and other operational characteristics of these systems remain largely unknown. Signification deviations from normal reactor heat removal mechanisms exist, because the heat pipes have fundamental heat removal limits due to sonic flow issues at low temperatures. This paper proposes an earlymore » prototypic test of a Heat Pipe Reactor (using existing 20% enriched nuclear fuel pins) to determine the operational characteristics of the HPR. The proposed design is similar in design to the HOMER and SAFE-300 HPR designs (Elliot, Lipinski, and Poston, 2003; Houts, et. al, 2003). However, this reactor uses existing UZrH fuel pins that are coupled to potassium heat pipes modules. The prototype reactor would be located in the Sandia Annular Core Research Reactor Facility where the fuel pins currently reside. The proposed reactor would use the heat pipes to transport the heat from the UZrH fuel pins to a water pool above the core, and the heat transport to the water pool would be controlled by adjusting the pressure and gas type within a small annulus around each heat pipe. The reactor would operate as a self-critical assembly at power levels up to 200 kWth. Because the nuclear heated HPR test uses existing fuel and because it would be performed in an existing facility with the appropriate safety authorization basis, the test could be performed rapidly and inexpensively. This approach makes it possible to validate the operation of a HPR and also measure the feedback mechanisms for a typical HPR design. A test of this nature would be the world's first operating Heat Pipe Reactor. This reactor is therefore called 'HPR-1'.« less

Environmental assessment of SP-100 ground engineering system test site: Hanford Site, Richland, Washington

NASA Astrophysics Data System (ADS)

1988-12-01

The US Department of Energy (DOE) proposes to modify an existing reactor containment building (decommissioned Plutonium Recycle Test Reactor (PRTR) 309 Building) to provide ground test capability for the prototype SP-100 reactor. The 309 Building (Figure 1.1) is located in the 300 Area on the Hanford Site in Washington State. The National Environmental Policy Act (NEPA) requires that Federal agencies assess the potential impacts that their actions may have on the environment. This Environmental Assessment describes the consideration given to environmental impacts during reactor concept and test site selection, examines the environmental effects of the DOE proposal to ground test the nuclear subsystem, describes alternatives to the proposed action, and examines radiological risks of potential SP-100 use in space.

Novel Materials for Prosthetic Liners

NASA Technical Reports Server (NTRS)

Ragolta, Carolina I.; Morford, Megan

2011-01-01

Existing materials for prosthetic liners tend to be thick and airtight, causing perspiration to accumulate inside the liner and potentially causing infection and injury that reduce quality of life. The purpose of this project was to examine the suitability of aerogel for prosthetic liner applications. Three tests were performed on several types of aerogel to assess the properties of each material. Moisture vapor permeability was tested by incubating four aerogel varieties with an artificial sweat solution at 37.0 C and less than 20% relative humidity for 24 hours. Two aerogel varieties were eliminated from the study due to difficulties in handling the material, and further testing proceeded with Pyrogel in 2.0 and 6.0 mm thicknesses. Force distribution was tested by compressing samples under a load of 4448 N at a rate of 2.5 mm/min. Biofilm formation was tested in a high-shear CDC Biofilm Reactor. Results showed that 2.0 mm Pyrogel blanket allowed 55.7 plus or minus 28.7% of an artificial sweat solution to transpire, and 35.5 plus or minus 27.8% transpired through 6.0 mm Pyrogel blanket. Samples also outperformed the load-bearing capabilities of existing liner materials. No statistically significant difference was found between the two Pyrogel thicknesses for either moisture vapor permeability or force distribution. In addition, biofilm formation results showed no change between the two Pyrogel thicknesses. The breathability and load bearing properties of aerogel make it a suitable material for application to prosthetic liners.

An evaluation of alloys and coatings for use in automobile thermal reactors

NASA Technical Reports Server (NTRS)

Blankenship, C. P.; Oldrieve, R. E.

1974-01-01

Several candidate alloys and coatings were evaluated for use in automobile thermal reactors. Full-size reactors of the candidate materials were analyzed in cyclic engine dynamometer tests with peak temperature of 1900 F (1040 C). Two developmental ferritic iron alloys GE1541 and NASA-18T - exhibited the best overall performance lasting at least 60% of the life of the test engine. Four of the alloys evaluated warrant consideration for reactor use. They include GE1541, Armco 18 SR, NASA-18T, and Inconel 601. None of the commercial coating substrate combinations evaluated warrant consideration for reactor use.-

Evaluation of alloys and coatings for use in automobile thermal reactors

NASA Technical Reports Server (NTRS)

Blankenship, C. P.; Oldrieve, R. E.

1974-01-01

Several candidate alloys and coatings were evaluated for use in automobile thermal reactors. Full-size reactors of the candidate materials were evaluated in cyclic engine dynamometer tests with a peak temperature of 1040 C (1900 F). Two developmental ferritic-iron alloys, GE-1541 and NASA-18T, exhibited the best overall performance by lasting at least 60 percent of the life of test engine. Four of the alloys evaluated warrant consideration for reactor use. They are GE-1541, Armco 18 SR, NASA-18T, and Inconel 601. None of the commercial coating substrate combinations evaluated warrant consideration for reactor use.

The effect of catalyst length and downstream reactor distance on catalytic combustor performance

NASA Technical Reports Server (NTRS)

Anderson, D.

1980-01-01

A study was made to determine the effects on catalytic combustor performance which resulted from independently varying the length of a catalytic reactor and the length available for gas-phase reactions downstream of the catalyst. Monolithic combustion catalysts from three manufacturers were tested in a combustion test rig with no. 2 diesel fuel. Catalytic reactor lengths of 2.5 and 5.4 cm, and downstream gas-phase reaction distances of 7.3, 12.4, 17.5, and 22.5 cm were evaluated. Measurements of carbon monoxide, unburned hydrocarbons, nitrogen oxides, and pressure drop were made. The catalytic-reactor pressure drop was less than 1 percent of the upstream total pressure for all test configurations and test conditions. Nitrogen oxides and unburned hydrocarbons emissions were less than 0.25 g NO2/kg fuel and 0.6 g HC/kg fuel, respectively. The minimum operating temperature (defined as the adiabatic combustion temperature required to obtain carbon monoxide emissions below a reference level of 13.6 g CO/kg fuel) ranged from 1230 K to 1500 K for the various conditions and configurations tested. The minimum operating temperature decreased with increasing total (catalytic-reactor-plus-downstream-gas-phase-reactor-zone) residence time but was independent of the relative times spent in each region when the catalytic-reactor residence time was greater than or equal to 1.4 ms.

EAST FACE OF REACTOR BASE. COMING TOWARD CAMERA IS EXCAVATION ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

EAST FACE OF REACTOR BASE. COMING TOWARD CAMERA IS EXCAVATION FOR MTR CANAL. CAISSONS FLANK EACH SIDE. COUNTERFORT (SUPPORT PERPENDICULAR TO WHAT WILL BE THE LONG WALL OF THE CANAL) RESTS ATOP LEFT CAISSON. IN LOWER PART OF VIEW, DRILLERS PREPARE TRENCHES FOR SUPPORT BEAMS THAT WILL LIE BENEATH CANAL FLOOR. INL NEGATIVE NO. 739. Unknown Photographer, 10/6/1950 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR, TRA642. ON GROUND FLOOR. THE 60TON ETR REACTOR VESSEL ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR, TRA-642. ON GROUND FLOOR. THE 60-TON ETR REACTOR VESSEL IS DROPPED INTO PLACE OVER PIT. KAISER USED TWO MULTI-BLOCK DRUM PULLEYS WITH A COMBINED CAPACITY OF 100 TONS AND A 100-TON DRUM HOIST. THE VESSEL WAS 35 1/2 FEET LONG. INL NEGATIVE NO. 56-4149. R.G. Larsen, Photographer, 12/18/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

LOFT. Reactor arrives at containment building (TAN650), now being pushed ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

LOFT. Reactor arrives at containment building (TAN-650), now being pushed by locomotive. Camera facing northerly. Note "Hello Dolly" and "PWR MTA No. 1" (pressurized water reactor mobile test assembly) signs. Date: 1973. INEEL negative no. 73-3710 - Idaho National Engineering Laboratory, Test Area North, Scoville, Butte County, ID

ETR BUILDING, TRA642, INTERIOR. CONSOLE FLOOR, SOUTH HALF. SOUTH SIDE ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR BUILDING, TRA-642, INTERIOR. CONSOLE FLOOR, SOUTH HALF. SOUTH SIDE OF ETR REACTOR, CAMERA FACING NORTH. CABINET CONTAINING "NUCLEAR INSTRUMENT SYSTEMS" IS RESTRICTED. INL NEGATIVE NO. HD46-18-4. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

FAST CHOPPER BUILDING, TRA665. CONTEXTUAL VIEW: CHOPPER BUILDING IN CENTER. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

FAST CHOPPER BUILDING, TRA-665. CONTEXTUAL VIEW: CHOPPER BUILDING IN CENTER. MTR REACTOR SERVICES BUILDING,TRA-635, TO LEFT; MTR BUILDING TO RIGHT. CAMERA FACING WEST. INL NEGATIVE NO. HD42-1. Mike Crane, Photographer, 3/2004 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

10 CFR 140.11 - Amounts of financial protection for certain reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

...,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding ten kilowatts; (2) In the amount of $1,500,000 for each nuclear reactor he is authorized to operate at... amount of $2,500,000 for each nuclear reactor other than a testing reactor or a reactor licensed under...

10 CFR 140.11 - Amounts of financial protection for certain reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

...,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding ten kilowatts; (2) In the amount of $1,500,000 for each nuclear reactor he is authorized to operate at... amount of $2,500,000 for each nuclear reactor other than a testing reactor or a reactor licensed under...

10 CFR 140.11 - Amounts of financial protection for certain reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

...,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding ten kilowatts; (2) In the amount of $1,500,000 for each nuclear reactor he is authorized to operate at... amount of $2,500,000 for each nuclear reactor other than a testing reactor or a reactor licensed under...

SPERT I DESTRUCTIVE TEST PROGRAM SAFETY ANALYSIS REPORT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Spano, A.H.; Miller, R.W.

1962-06-15

The water-moderated core used for destructive experiments is mounted in the Spent I open-type reactor vessel, which has no provision for pressurization or forced coolant flow. The core is an array of highly enriched aluminum clad, plate-type fuel assemblies, using four bladetype, gang-operated control rods. Reactor transients are initiated at ambient temperature by step-insentions of reactivity, using a control rod which can be quickly ejected from the core. Following an initial series of static measurements to determine the basic- reactor properties of the test core, a series of nondestructive, self-limiting power excursion tests was performed, which covered a reactor periodmore » range down to the point where minor fuel plate damage first occurred -approximately for a 10- msec period test. These tests provided power, temperature, and pressure data. Additional kinetic teste in the period region between 10 and 5 msec were completed to explore the region of limited core damage. Fuel plate damage results included plate distortion, cladding cracking, and fuel melting. These exploratory tests were valuable in revealing unexpected changes in the dependence of pressure, temperature, burst energy, and burst shape parameters on reactor period, although the dependence of peak power on reactor period was not significantly changed. An evaluation of hazards involved in conducting the 2- msec test, based on pessimistic assumptions regarding fission product release and weather conditions, indicates that with the procedural controls normally exercised in the conduct of any transient test at Spent and the special controls to be in effect during the destructive test series, no significant hazard to personnel or to the general public will be obtained. All nuclear operation is conducted remotely approximately 1/2 mile from the reactor building. Discussion is also given of the supervision and control of personnel during and after each destructive test, and of the plans for re-entry, cleanup, and restoration of the facility. (auth)« less

Analysis of space reactor system components: Investigation through simulation and non-nuclear testing

NASA Astrophysics Data System (ADS)

Bragg-Sitton, Shannon M.

The use of fission energy in space power and propulsion systems offers considerable advantages over chemical propulsion. Fission provides over six orders of magnitude higher energy density, which translates to higher vehicle specific impulse and lower specific mass. These characteristics enable ambitious space exploration missions. The natural space radiation environment provides an external source of protons and high energy, high Z particles that can result in the production of secondary neutrons through interactions in reactor structures. Applying the approximate proton source in geosynchronous orbit during a solar particle event, investigation using MCNPX 2.5.b for proton transport through the SAFE-400 heat pipe cooled reactor indicates an incoming secondary neutron current of (1.16 +/- 0.03) x 107 n/s at the core-reflector interface. This neutron current may affect reactor operation during low power maneuvers (e.g., start-up) and may provide a sufficient reactor start-up source. It is important that a reactor control system be designed to automatically adjust to changes in reactor power levels, maintaining nominal operation without user intervention. A robust, autonomous control system is developed and analyzed for application during reactor start-up, accounting for fluctuations in the radiation environment that result from changes in vehicle location or to temporal variations in the radiation field. Development of a nuclear reactor for space applications requires a significant amount of testing prior to deployment of a flight unit. High confidence in fission system performance can be obtained through relatively inexpensive non-nuclear tests performed in relevant environments, with the heat from nuclear fission simulated using electric resistance heaters. A series of non-nuclear experiments was performed to characterize various aspects of reactor operation. This work includes measurement of reactor core deformation due to material thermal expansion and implementation of a virtual reactivity feedback control loop; testing and thermal hydraulic characterization of the coolant flow paths for two space reactor concepts; and analysis of heat pipe operation during start-up and steady state operation.

Biological hydrogen production by Clostridium acetobutylicum in an unsaturated flow reactor.

PubMed

Zhang, Husen; Bruns, Mary Ann; Logan, Bruce E

2006-02-01

A mesophilic unsaturated flow (trickle bed) reactor was designed and tested for H2 production via fermentation of glucose. The reactor consisted of a column packed with glass beads and inoculated with a pure culture (Clostridium acetobutylicum ATCC 824). A defined medium containing glucose was fed at a flow rate of 1.6 mL/min (0.096 L/h) into the capped reactor, producing a hydraulic retention time of 2.1 min. Gas-phase H2 concentrations were constant, averaging 74 +/- 3% for all conditions tested. H2 production rates increased from 89 to 220 mL/hL of reactor when influent glucose concentrations were varied from 1.0 to 10.5 g/L. Specific H2 production rate ranged from 680 to 1270 mL/g glucose per liter of reactor (total volume). The H2 yield was 15-27%, based on a theoretical limit by fermentation of 4 moles of H2 from 1 mole of glucose. The major fermentation by-products in the liquid effluent were acetate and butyrate. The reactor rapidly (within 60-72 h) became clogged with biomass, requiring manual cleaning of the system. In order to make long-term operation of the reactor feasible, biofilm accumulation in the reactor will need to be controlled through some process such as backwashing. These tests using an unsaturated flow reactor demonstrate the feasibility of the process to produce high H2 gas concentrations in a trickle-bed type of reactor. A likely application of this reactor technology could be H2 gas recovery from pre-treatment of high carbohydrate-containing wastewaters.

Reactor Testing and Qualification: Prioritized High-level Criticality Testing Needs

DOE Office of Scientific and Technical Information (OSTI.GOV)

S. Bragg-Sitton; J. Bess; J. Werner

2011-09-01

Researchers at the Idaho National Laboratory (INL) were tasked with reviewing possible criticality testing needs to support development of the fission surface power system reactor design. Reactor physics testing can provide significant information to aid in development of technologies associated with small, fast spectrum reactors that could be applied for non-terrestrial power systems, leading to eventual system qualification. Several studies have been conducted in recent years to assess the data and analyses required to design and build a space fission power system with high confidence that the system will perform as designed [Marcille, 2004a, 2004b; Weaver, 2007; Parry et al.,more » 2008]. This report will provide a summary of previous critical tests and physics measurements that are potentially applicable to the current reactor design (both those that have been benchmarked and those not yet benchmarked), summarize recent studies of potential nuclear testing needs for space reactor development and their applicability to the current baseline fission surface power (FSP) system design, and provide an overview of a suite of tests (separate effects, sub-critical or critical) that could fill in the information database to improve the accuracy of physics modeling efforts as the FSP design is refined. Some recommendations for tasks that could be completed in the near term are also included. Specific recommendations on critical test configurations will be reserved until after the sensitivity analyses being conducted by Los Alamos National Laboratory (LANL) are completed (due August 2011).« less

Assessment of the Technical Maturity of Generation IV Concepts for Test or Demonstration Reactor Applications, Revision 2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gougar, Hans David

2015-10-01

The United States Department of Energy (DOE) commissioned a study the suitability of different advanced reactor concepts to support materials irradiations (i.e. a test reactor) or to demonstrate an advanced power plant/fuel cycle concept (demonstration reactor). As part of the study, an assessment of the technical maturity of the individual concepts was undertaken to see which, if any, can support near-term deployment. A Working Group composed of the authors of this document performed the maturity assessment using the Technical Readiness Levels as defined in DOE’s Technology Readiness Guide . One representative design was selected for assessment from of each ofmore » the six Generation-IV reactor types: gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), molten salt reactor (MSR), supercritical water-cooled reactor (SCWR), sodium-cooled fast reactor (SFR), and very high temperature reactor (VHTR). Background information was obtained from previous detailed evaluations such as the Generation-IV Roadmap but other technical references were also used including consultations with concept proponents and subject matter experts. Outside of Generation IV activity in which the US is a party, non-U.S. experience or data sources were generally not factored into the evaluations as one cannot assume that this data is easily available or of sufficient quality to be used for licensing a US facility. The Working Group established the scope of the assessment (which systems and subsystems needed to be considered), adapted a specific technology readiness scale, and scored each system through discussions designed to achieve internal consistency across concepts. In general, the Working Group sought to determine which of the reactor options have sufficient maturity to serve either the test or demonstration reactor missions.« less

Fabrication and testing of a 4-node micro-pocket fission detector array for the Kansas State University TRIGA Mk. II research nuclear reactor

NASA Astrophysics Data System (ADS)

Reichenberger, Michael A.; Nichols, Daniel M.; Stevenson, Sarah R.; Swope, Tanner M.; Hilger, Caden W.; Unruh, Troy C.; McGregor, Douglas S.; Roberts, Jeremy A.

2017-08-01

Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Micro-Pocket Fission Detectors (MPFDs) have been fabricated and tested previously, but successful testing of these prior detectors was limited to single-node operation with specialized designs. Described in this work is a modular, four-node MPFD array fabricated and tested at Kansas State University (KSU). The four sensor nodes were equally spaced to span the length of the fuel-region of the KSU TRIGA Mk. II research nuclear reactor core. The encapsulated array was filled with argon gas, serving as an ionization medium in the small cavities of the MPFDs. The unified design improved device ruggedness and simplified construction over previous designs. A 0.315-in. (8-mm) penetration in the upper grid plate of the KSU TRIGA Mk. II research nuclear reactor was used to deploy the array between fuel elements in the core. The MPFD array was coupled to an electronic support system which has been developed to support pulse-mode operation. Neutron-induced pulses were observed on all four sensor channels. Stable device operation was confirmed by testing under steady-state reactor conditions. Each of the four sensors in the array responded to changes in reactor power between 10 kWth and full power (750 kWth). Reactor power transients were observed in real-time including positive transients with periods of 5, 15, and 30 s. Finally, manual reactor power oscillations were observed in real-time.

10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

Code of Federal Regulations, 2014 CFR

2014-01-01

... air test pressure and to assure they will be subjected to the post accident differential pressure... Table of Contents I. Introduction. II. Explanation of terms. III. Leakage test requirements. A. Type A test. B. Type B test. C. Type C test. D. Periodic retest schedule. IV. Special test requirements. A...

10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

Code of Federal Regulations, 2012 CFR

2012-01-01

... air test pressure and to assure they will be subjected to the post accident differential pressure... Table of Contents I. Introduction. II. Explanation of terms. III. Leakage test requirements. A. Type A test. B. Type B test. C. Type C test. D. Periodic retest schedule. IV. Special test requirements. A...

10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

Code of Federal Regulations, 2010 CFR

2010-01-01

... air test pressure and to assure they will be subjected to the post accident differential pressure... Table of Contents I. Introduction. II. Explanation of terms. III. Leakage test requirements. A. Type A test. B. Type B test. C. Type C test. D. Periodic retest schedule. IV. Special test requirements. A...

10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

Code of Federal Regulations, 2013 CFR

2013-01-01

... air test pressure and to assure they will be subjected to the post accident differential pressure... Table of Contents I. Introduction. II. Explanation of terms. III. Leakage test requirements. A. Type A test. B. Type B test. C. Type C test. D. Periodic retest schedule. IV. Special test requirements. A...

Features of postfailure fuel behavior in transient overpower and transient undercooled/overpower tests in the transient reactor test facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Doerner, R.C.; Bauer, T.H.; Morman, J.A.

Prototypic oxide fuel was subjected to simulated, fast reactor severe accident conditions in a series of in-pile tests in the Transient Reactor Test Facility reactor. Seven experiments were performed on fresh and previously irradiated oxide fuel pins under transient overpower and transient undercooled. overpower accident conditions. For each of the tests, fuel motions were observed by the hodoscope. Hodoscope data are correlated with coolant flow, pressure, and temperature data recorded by the loop instrumentation. Data were analyzed from the onset of initial failure to a final mass distribution at the end of the test. In this paper results of thesemore » analyses are compared to pre- and posttest accident calculations and to posttest metallographic accident calculations and to posttest metallographic examinations and computed tomographic reconstructions from neutron radiographs.« less

World Energy Data System (WENDS). Volume XI. Nuclear fission program summaries

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1979-06-01

Brief management and technical summaries of nuclear fission power programs are presented for nineteen countries. The programs include the following: fuel supply, resource recovery, enrichment, fuel fabrication, light water reactors, heavy water reactors, gas cooled reactors, breeder reactors, research and test reactors, spent fuel processing, waste management, and safety and environment. (JWR)

2007 international meeting on Reduced Enrichment for Research and Test Reactors (RERTR). Abstracts and available papers presented at the meeting

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

2008-07-15

The Meeting papers discuss research and test reactor fuel performance, manufacturing and testing. Some of the main topics are: conversion from HEU to LEU in different reactors and corresponding problems and activities; flux performance and core lifetime analysis with HEU and LEU fuels; physics and safety characteristics; measurement of gamma field parameters in core with LEU fuel; nondestructive analysis of RERTR fuel; thermal hydraulic analysis; fuel interactions; transient analyses and thermal hydraulics for HEU and LEU cores; microstructure research reactor fuels; post irradiation analysis and performance; computer codes and other related problems.

MONTE CARLO SIMULATIONS OF PERIODIC PULSED REACTOR WITH MOVING GEOMETRY PARTS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cao, Yan; Gohar, Yousry

2015-11-01

In a periodic pulsed reactor, the reactor state varies periodically from slightly subcritical to slightly prompt supercritical for producing periodic power pulses. Such periodic state change is accomplished by a periodic movement of specific reactor parts, such as control rods or reflector sections. The analysis of such reactor is difficult to perform with the current reactor physics computer programs. Based on past experience, the utilization of the point kinetics approximations gives considerable errors in predicting the magnitude and the shape of the power pulse if the reactor has significantly different neutron life times in different zones. To accurately simulate themore » dynamics of this type of reactor, a Monte Carlo procedure using the transfer function TRCL/TR of the MCNP/MCNPX computer programs is utilized to model the movable reactor parts. In this paper, two algorithms simulating the geometry part movements during a neutron history tracking have been developed. Several test cases have been developed to evaluate these procedures. The numerical test cases have shown that the developed algorithms can be utilized to simulate the reactor dynamics with movable geometry parts.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

Progress is reported on fundamental research in: crystal physics, reactions at metal surfaces, spectroscopy of ionic media, structure of metals, theory of alloying, physical properties, sintering, deformation of crystalline solids, x ray diffraction, metallurgy of superconducting materials, and electron microscope studies. Long-randge applied research studies were conducted for: zirconium metallurgy, materials compatibility, solid reactions, fuel element development, mechanical properties, non-destructive testing, and high-temperature materials. Reactor development support work was carried out for: gas-cooled reactor program, molten-salt reactor, high-flux isotope reactor, space-power program, thorium-utilization program, advanced-test reactor, Army Package Power Reactor, Enrico Fermi fast-breeder reactor, and water desalination program. Other programmore » activities, for which research was conducted, included: thermonuclear project, transuraniunn program, and post-irradiation examination laboratory. Separate abstracts were prepared for 30 sections of the report. (B.O.G.)« less

MTR, TRA603. FIRST FLOOR PLAN. REACTOR AT CENTER. TWENTYMETER CHOPPER ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

MTR, TRA-603. FIRST FLOOR PLAN. REACTOR AT CENTER. TWENTY-METER CHOPPER HOUSE. COFFIN TURNING ROLLS. REMOVABLE PANEL OVER CANAL ON EAST SIDE. NEW PLUG STORAGE ACCESS. DOOR SCHEDULE INDICATES STEEL (FOR VAULT), WIRE MESH, AND HOLLOW METAL TYPES. STORAGE AND ISSUE ROOM. SAFETY SHOWERS. DOORWAY TO WING, TRA-604. BLAW-KNOX 3150-803-2, 7/1950. INL INDEX NO. 531-0603-00-098-100561, REV. 10. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1982

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E.

1983-06-01

During 1982 the High-Temperature Gas-Cooled Reactor (HTGR) Technology Program at Oak Ridge National Laboratory (ORNL) continued to develop experimental data required for the design and licensing of cogeneration HTGRs. The program involves fuels and materials development (including metals, graphite, ceramic, and concrete materials), HTGR chemistry studies, structural component development and testing, reactor physics and shielding studies, performance testing of the reactor core support structure, and HTGR application and evaluation studies.

REACTOR SERVICE BUILDING, TRA635, INTERIOR. CAMERA FACES NORTHWEST TOWARDS INTERIOR ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

REACTOR SERVICE BUILDING, TRA-635, INTERIOR. CAMERA FACES NORTHWEST TOWARDS INTERIOR WALL ENCLOSING STORAGE AND OFFICE SPACE ALONG THE WEST SIDE. AT RIGHT EDGE IS DOOR TO MTR BUILDING. FROM RIGHT TO LEFT, SPACE WAS PLANNED FOR A LOCKER ROOM, MTR ISSUE ROOM, AND STORAGE AREAS AND RELATED OFFICES. NOTE SECOND "MEZZANINE" FLOOR ABOVE. INL NEGATIVE NO. 10227. Unknown Photographer, 3/23/1954 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

NASA-EPA automotive thermal reactor technology program

NASA Technical Reports Server (NTRS)

Blankenship, C. P.; Hibbard, R. R.

1972-01-01

The status of the NASA-EPA automotive thermal reactor technology program is summarized. This program is concerned primarily with materials evaluation, reactor design, and combustion kinetics. From engine dynamometer tests of candidate metals and coatings, two ferritic iron alloys (GE 1541 and Armco 18-SR) and a nickel-base alloy (Inconel 601) offer promise for reactor use. None of the coatings evaluated warrant further consideration. Development studies on a ceramic thermal reactor appear promising based on initial vehicle road tests. A chemical kinetic study has shown that gas temperatures of at least 900 K to 1000 K are required for the effective cleanup of carbon monoxide and hydrocarbons, but that higher temperatures require shorter combustion times and thus may permit smaller reactors.

75 FR 34219 - Revision of Fee Schedules; Fee Recovery for FY 2010

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-16

....8 $6.3 $7.5 Spent Fuel Storage/Reactor Decommissioning..... -- -- 2.7 0.2 0.2 Test and Research... 2009 fee is also shown for comparative purposes. Table V--Rebaselined Annual Fees FY2009 Annual FY 2010... Decommissioning Test and Research Reactors (Non-power 87,600 81,700 Reactors) High Enriched Uranium Fuel Facility...

ETR, TRA642. ON GROUND FLOOR. WITH OUTER THERMAL RING IN ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR, TRA-642. ON GROUND FLOOR. WITH OUTER THERMAL RING IN PLACE AND CONDUIT PRESERVED, HIGH-DENSITY CONCRETE IS PLACED BETWEEN THE THERMAL RING AND THE OUTER REACTOR FORM. INL NEGATIVE NO. 56-2400. Jack L. Anderson, Photographer, 6/10/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Apparatus and method for simultaneous recovery of hydrogen from water and from hydrocarbons

DOEpatents

Willms, R. Scott; Birdsell, Stephen A.

2000-01-01

Apparatus and method for simultaneous recovery of hydrogen from water and from hydrocarbon feed material. The feed material is caused to flow over a heated catalyst which fosters the water-gas shift reaction (H.sub.2 O+COH.sub.2 +CO.sub.2) and the methane steam reforming reaction (CH.sub.4 +H.sub.2 O3 H.sub.2 +CO). Both of these reactions proceed only to partial completion. However, by use of a Pd/Ag membrane which is exclusively permeable to hydrogen isotopes in the vicinity of the above reactions and by maintaining a vacuum on the permeate side of the membrane, product hydrogen isotopes are removed and the reactions are caused to proceed further toward completion. A two-stage palladium membrane reactor was tested with a feed composition of 28% CQ.sub.4, 35% Q.sub.2 O (where Q=H, D, or T), and 31% Ar in 31 hours of continuous operation during which 4.5 g of tritium were processed. Decontamination factors were found to increase with decreasing inlet rate. The first stage was observed to have a decontamination factor of approximately 200, while the second stage had a decontamination factor of 2.9.times.10.sup.6. The overall decontamination factor was 5.8.times.10.sup.8. When a Pt/.alpha.-Al.sub.2 O.sub.3 catalyst is employed, decoking could be performed without catalyst degradation. However, by adjusting the carbon to oxygen ratio of the feed material with the addition of oxygen, coking could be altogether avoided.

Corrosion Evaluation of RERTR Uranium Molybdenum Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

A K Wertsching

2012-09-01

As part of the National Nuclear Security Agency (NNSA) mandate to replace the use of highly enriched uranium (HEU) fuel for low enriched uranium (LEU) fuel, research into the development of LEU fuel for research reactors has been active since the late 1970’s. Originally referred to as the Reduced Enrichment for Research and Test Reactor (RERTR) program the new effort named Global Threat Reduction Initiative (GTRI) is nearing the goal of replacing the standard aluminum clad dispersion highly enriched uranium aluminide fuel with a new LEU fuel. The five domestic high performance research reactors undergoing this conversion are High Fluxmore » Isotope reactor (HFIR), Advanced Test Reactor (ATR), National Institute of Standards and Technology (NIST) Reactor, Missouri University Research Reactor (MURR) and the Massachusetts Institute of Technology Reactor II (MITR-II). The design of these reactors requires a higher neutron flux than other international research reactors, which to this point has posed unique challenges in the design and development of the new mandated LEU fuel. The new design utilizes a monolithic fuel configuration in order to obtain sufficient 235U within the LEU stoichoimetry to maintain the fission reaction within the domestic test reactors. The change from uranium aluminide dispersion fuel type to uranium molybdenum (UMo) monolithic configuration requires examination of possible corrosion issues associated with the new fuel meat. A focused analysis of the UMo fuel under potential corrosion conditions, within the ATR and under aqueous storage indicates a slow and predictable corrosion rate. Additional corrosion testing is recommended for the highest burn-up fuels to confirm observed corrosion rate trends. This corrosion analysis will focus only on the UMo fuel and will address corrosion of ancillary components such as cladding only in terms of how it affects the fuel. The calculations and corrosion scenarios are weighted with a conservative bias to provide additional confidence with the results. The actual corrosion rates of UMo fuel is very likely to be lower than assumed within this report which can be confirmed with additional testing.« less

HISTORICAL AMERICAN ENGINEERING RECORD - IDAHO NATIONAL ENGINEERING AND ENVIRONMENTAL LABORATORY, TEST AREA NORTH, HAER NO. ID-33-E

DOE Office of Scientific and Technical Information (OSTI.GOV)

Susan Stacy; Hollie K. Gilbert

2005-02-01

Test Area North (TAN) was a site of the Aircraft Nuclear Propulsion (ANP) Project of the U.S. Air Force and the Atomic Energy Commission. Its Cold War mission was to develop a turbojet bomber propelled by nuclear power. The project was part of an arms race. Test activities took place in five areas at TAN. The Assembly & Maintenance area was a shop and hot cell complex. Nuclear tests ran at the Initial Engine Test area. Low-power test reactors operated at a third cluster. The fourth area was for Administration. A Flight Engine Test facility (hangar) was built to housemore » the anticipated nuclear-powered aircraft. Experiments between 1955-1961 proved that a nuclear reactor could power a jet engine, but President John F. Kennedy canceled the project in March 1961. ANP facilities were adapted for new reactor projects, the most important of which were Loss of Fluid Tests (LOFT), part of an international safety program for commercial power reactors. Other projects included NASA's Systems for Nuclear Auxiliary Power and storage of Three Mile Island meltdown debris. National missions for TAN in reactor research and safety research have expired; demolition of historic TAN buildings is underway.« less

Single Channel Testing for Characterization of the Direct Gas Cooled Reactor and the SAFE-100 Heat Exchanger

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bragg-Sitton, S.M.; Propulsion Research Center, NASA Marshall Space Flight Center, Huntsville, AL 35812; Kapernick, R.

2004-02-04

Experiments have been designed to characterize the coolant gas flow in two space reactor concepts that are currently under investigation by NASA Marshall Space Flight Center and Los Alamos National Laboratory: the direct-drive gas-cooled reactor (DDG) and the SAFE-100 heatpipe-cooled reactor (HPR). For the DDG concept, initial tests have been completed to measure pressure drop versus flow rate for a prototypic core flow channel, with gas exiting to atmospheric pressure conditions. The experimental results of the completed DDG tests presented in this paper validate the predicted results to within a reasonable margin of error. These tests have resulted in amore » re-design of the flow annulus to reduce the pressure drop. Subsequent tests will be conducted with the re-designed flow channel and with the outlet pressure held at 150 psi (1 MPa). Design of a similar test for a nominal flow channel in the HPR heat exchanger (HPR-HX) has been completed and hardware is currently being assembled for testing this channel at 150 psi. When completed, these test programs will provide the data necessary to validate calculated flow performance for these reactor concepts (pressure drop and film temperature rise)« less

Effects of glucose on the performance of enhanced biological phosphorus removal activated sludge enriched with acetate.

PubMed

Gebremariam, Seyoum Yami; Beutel, Marc W; Christian, David; Hess, Thomas F

2012-10-01

The effects of glucose on enhanced biological phosphorus removal (EBPR) activated sludge enriched with acetate was investigated using sequencing batch reactors. A glucose/acetate mixture was serially added to the test reactor in ratios of 25/75%, 50/50%, and 75/25% and the EBPR activity was compared to the control reactor fed with 100% acetate. P removal increased at a statistically significant level to a near-complete in the test reactor when the mixture increased to 50/50%. However, EBPR deteriorated when the glucose/acetate mixture increased to 75/25% in the test reactor and when the control reactor abruptly switched to 100% glucose. These results, in contrast to the EBPR conventional wisdom, suggest that the addition of glucose at moderate levels in wastewaters does not impede and may enhance EBPR, and that glucose waste products should be explored as an economical sustainable alternative when COD enhancement of EBPR is needed. Copyright © 2012 Elsevier Ltd. All rights reserved.

Flat-plate collector research area: Silicon material task

NASA Technical Reports Server (NTRS)

Lutwack, R.

1982-01-01

Silane decomposition in a fluidized-bed reactor (FBR) process development unit (PDU) to make semiconductor-grade Si is reviewed. The PDU was modified by installation of a new heating system to provide the required temperature profile and better control, and testing was resumed. A process for making trichlorosilane by the hydrochlorination of metallurgical-grade Si and silicon tetrachloride is reported. Fabrication and installation of the test system employing a new 2-in.-dia reactor was completed. A process that converts trichlorosilane to dichlorosilane (DCS), which is reduced by hydrogen to make Si by a chemical vapor deposition step in a Siemens-type reactor is described. Testing of the DCS PDU integraled with Si deposition reactors continued. Experiments in a 2-in.-dia reactor to define the operating window and to investigate the Si deposition kinetics were completed.

Multiple Irradiation Capsule Experiment (MICE)-3B Irradiation Test of Space Fuel Specimens in the Advanced Test Reactor (ATR) - Close Out Documentation for Naval Reactors (NR) Information

DOE Office of Scientific and Technical Information (OSTI.GOV)

M. Chen; CM Regan; D. Noe

2006-01-09

Few data exist for UO{sub 2} or UN within the notional design space for the Prometheus-1 reactor (low fission rate, high temperature, long duration). As such, basic testing is required to validate predictions (and in some cases determine) performance aspects of these fuels. Therefore, the MICE-3B test of UO{sub 2} pellets was designed to provide data on gas release, unrestrained swelling, and restrained swelling at the upper range of fission rates expected for a space reactor. These data would be compared with model predictions and used to determine adequacy of a space reactor design basis relative to fission gas releasemore » and swelling of UO{sub 2} fuel and to assess potential pellet-clad interactions. A primary goal of an irradiation test for UN fuel was to assess performance issues currently associated with this fuel type such as gas release, swelling and transient performance. Information learned from this effort may have enabled use of UN fuel for future applications.« less

Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

2009-01-01

A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, OH. This is a closed-cycle system that incorporates an electrically heated reactor core module, turbo alternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

2010-01-01

A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, Ohio. This is a closed-cycle system that incorporates an electrically heated reactor core module, turboalternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

Assessment of Nuclear Fuels using Radiographic Thickness Measurement Method

DOE Office of Scientific and Technical Information (OSTI.GOV)

Muhammad Abir; Fahima Islam; Hyoung Koo Lee

2014-11-01

The Convert branch of the National Nuclear Security Administration (NNSA) Global Threat Reduction Initiative (GTRI) focuses on the development of high uranium density fuels for research and test reactors for nonproliferation. This fuel is aimed to convert low density high enriched uranium (HEU) based fuel to high density low enriched uranium (LEU) based fuel for high performance research reactors (HPRR). There are five U.S. reactors that fall under the HPRR category, including: the Massachusetts Institute of Technology Reactor (MITR), the National Bureau of Standards Reactor (NBSR), the Missouri University Research Reactor (UMRR), the Advanced Test Reactor (ATR), and the Highmore » Flux Isotope Reactor (HFIR). U-Mo alloy fuel phase in the form of either monolithic or dispersion foil type fuels, such as ATR Full-size In center flux trap Position (AFIP) and Reduced Enrichment for Research and Test Reactor (RERTR), are being designed for this purpose. The fabrication process1 of RERTR is susceptible to introducing a variety of fuel defects. A dependable quality control method is required during fabrication of RERTR miniplates to maintain the allowable design tolerances, therefore evaluating and analytically verifying the fabricated miniplates for maintaining quality standards as well as safety. The purpose of this work is to analyze the thickness of the fabricated RERTR-12 miniplates using non-destructive technique to meet the fuel plate specification for RERTR fuel to be used in the ATR.« less

Advanced Thermal Simulator Testing: Thermal Analysis and Test Results

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Dickens, Ricky; Dixon, David; Reid, Robert; Adams, Mike; Davis, Joe

2008-01-01

Work at the NASA Marshall Space Flight Center seeks to develop high fidelity, electrically heated thermal simulators that represent fuel elements in a nuclear reactor design to support non-nuclear testing applicable to the development of a space nuclear power or propulsion system. Comparison between the fuel pins and thermal simulators is made at the outer fuel clad surface, which corresponds to the outer sheath surface in the thermal simulator. The thermal simulators that are currently being tested correspond to a SNAP derivative reactor design that could be applied for Lunar surface power. These simulators are designed to meet the geometric and power requirements of a proposed surface power reactor design, accommodate testing of various axial power profiles, and incorporate imbedded instrumentation. This paper reports the results of thermal simulator analysis and testing in a bare element configuration, which does not incorporate active heat removal, and testing in a water-cooled calorimeter designed to mimic the heat removal that would be experienced in a reactor core.

Caesium-rich micro-particles: A window into the meltdown events at the Fukushima Daiichi Nuclear Power Plant

PubMed Central

Furuki, Genki; Imoto, Junpei; Ochiai, Asumi; Yamasaki, Shinya; Nanba, Kenji; Ohnuki, Toshihiko; Grambow, Bernd; Ewing, Rodney C.; Utsunomiya, Satoshi

2017-01-01

The nuclear disaster at the Fukushima Daiichi Nuclear Power Plant (FDNPP) in March 2011 caused partial meltdowns of three reactors. During the meltdowns, a type of condensed particle, a caesium-rich micro-particle (CsMP), formed inside the reactors via unknown processes. Here we report the chemical and physical processes of CsMP formation inside the reactors during the meltdowns based on atomic-resolution electron microscopy of CsMPs discovered near the FDNPP. All of the CsMPs (with sizes of 2.0–3.4 μm) comprise SiO2 glass matrices and ~10-nm-sized Zn–Fe-oxide nanoparticles associated with a wide range of Cs concentrations (1.1–19 wt% Cs as Cs2O). Trace amounts of U are also associated with the Zn–Fe oxides. The nano-texture in the CsMPs records multiple reaction-process steps during meltdown in the severe FDNPP accident: Melted fuel (molten core)-concrete interactions (MCCIs), incorporating various airborne fission product nanoparticles, including CsOH and CsCl, proceeded via SiO2 condensation over aggregates of Zn-Fe oxide nanoparticles originating from the failure of the reactor pressure vessels. Still, CsMPs provide a mechanism by which volatile and low-volatility radionuclides such as U can reach the environment and should be considered in the migration model of Cs and radionuclides in the current environment surrounding the FDNPP. PMID:28198440

Caesium-rich micro-particles: A window into the meltdown events at the Fukushima Daiichi Nuclear Power Plant

NASA Astrophysics Data System (ADS)

Furuki, Genki; Imoto, Junpei; Ochiai, Asumi; Yamasaki, Shinya; Nanba, Kenji; Ohnuki, Toshihiko; Grambow, Bernd; Ewing, Rodney C.; Utsunomiya, Satoshi

2017-02-01

The nuclear disaster at the Fukushima Daiichi Nuclear Power Plant (FDNPP) in March 2011 caused partial meltdowns of three reactors. During the meltdowns, a type of condensed particle, a caesium-rich micro-particle (CsMP), formed inside the reactors via unknown processes. Here we report the chemical and physical processes of CsMP formation inside the reactors during the meltdowns based on atomic-resolution electron microscopy of CsMPs discovered near the FDNPP. All of the CsMPs (with sizes of 2.0-3.4 μm) comprise SiO2 glass matrices and ~10-nm-sized Zn-Fe-oxide nanoparticles associated with a wide range of Cs concentrations (1.1-19 wt% Cs as Cs2O). Trace amounts of U are also associated with the Zn-Fe oxides. The nano-texture in the CsMPs records multiple reaction-process steps during meltdown in the severe FDNPP accident: Melted fuel (molten core)-concrete interactions (MCCIs), incorporating various airborne fission product nanoparticles, including CsOH and CsCl, proceeded via SiO2 condensation over aggregates of Zn-Fe oxide nanoparticles originating from the failure of the reactor pressure vessels. Still, CsMPs provide a mechanism by which volatile and low-volatility radionuclides such as U can reach the environment and should be considered in the migration model of Cs and radionuclides in the current environment surrounding the FDNPP.

Radiation Effects in Fission and Fusion Reactors

NASA Astrophysics Data System (ADS)

Odette, G. Robert; Wirth, Brian D.

Since the prediction of "Wigner disease" [1] and the subsequent observation of anisotropic growth of the graphite used in the Chicago Pile, the effects of radiation on materials has been an important technological concern. The broad field of radiation effects impacts many critical advanced technologies, ranging from semiconductor processing to severe materials degradation in nuclear reactor environments. Radiation effects also occur in many natural environments, ranging from deep space to inside the Earth's crust. As selected examples that involve many basic phenomena that cross-cut and illustrate the broader impacts of radiation exposure on materials, this article focuses on modeling microstructural changes in iron-based ferritic alloys under high-energy neutron irradiation relevant to light water fission reactor pressure vessels. We also touch briefly on radiation effects in structural alloys for fusion reactor first wall and blanket structures; in this case the focus is on modeling the evolution of self-interstitial atom clusters and dislocation loops. Note, since even the narrower topic of structural materials for nuclear energy applications encompass a vast literature dating from 1942, the references included in this article are primarily limited to these two narrower subjects. Thus, the references cited here are presented as examples, rather than comprehensive bibliographies. However, the interested reader is referred to proceedings of continuing symposia series that have been sponsored by several organizations, several monographs [2-4] and key journals (e.g., Journal of Nuclear Materials, Radiation Effects and Defects in Solids).

Caesium-rich micro-particles: A window into the meltdown events at the Fukushima Daiichi Nuclear Power Plant.

PubMed

Furuki, Genki; Imoto, Junpei; Ochiai, Asumi; Yamasaki, Shinya; Nanba, Kenji; Ohnuki, Toshihiko; Grambow, Bernd; Ewing, Rodney C; Utsunomiya, Satoshi

2017-02-15

The nuclear disaster at the Fukushima Daiichi Nuclear Power Plant (FDNPP) in March 2011 caused partial meltdowns of three reactors. During the meltdowns, a type of condensed particle, a caesium-rich micro-particle (CsMP), formed inside the reactors via unknown processes. Here we report the chemical and physical processes of CsMP formation inside the reactors during the meltdowns based on atomic-resolution electron microscopy of CsMPs discovered near the FDNPP. All of the CsMPs (with sizes of 2.0-3.4 μm) comprise SiO 2 glass matrices and ~10-nm-sized Zn-Fe-oxide nanoparticles associated with a wide range of Cs concentrations (1.1-19 wt% Cs as Cs 2 O). Trace amounts of U are also associated with the Zn-Fe oxides. The nano-texture in the CsMPs records multiple reaction-process steps during meltdown in the severe FDNPP accident: Melted fuel (molten core)-concrete interactions (MCCIs), incorporating various airborne fission product nanoparticles, including CsOH and CsCl, proceeded via SiO 2 condensation over aggregates of Zn-Fe oxide nanoparticles originating from the failure of the reactor pressure vessels. Still, CsMPs provide a mechanism by which volatile and low-volatility radionuclides such as U can reach the environment and should be considered in the migration model of Cs and radionuclides in the current environment surrounding the FDNPP.

A novel plant protection strategy for transient reactors

NASA Astrophysics Data System (ADS)

Bhattacharyya, Samit K.; Lipinski, Walter C.; Hanan, Nelson A.

A novel plant protection system designed for use in the TREAT Upgrade (TU) reactor is described. The TU reactor is designed for controlled transient operation in the testing of reactor fuel behavior under simulated reactor accident conditions. Safe operation of the reactor is of paramount importance and the Plant Protection System (PPS) had to be designed to exacting requirements. Researchers believe that the strategy developed for the TU has potential application to the multimegawatt space reactors and represents the state of the art in terrestrial transient reactor protection systems.

NEUTRONIC REACTOR SHIELDING

DOEpatents

Borst, L.B.

1961-07-11

A special hydrogenous concrete shielding for reactors is described. In addition to Portland cement and water, the concrete essentially comprises 30 to 60% by weight barytes aggregate for enhanced attenuation of fast neutrons. The biological shields of AEC's Oak Ridge Graphite Reactor and Materials Testing Reactor are particular embodiments.

5 CFR 5801.102 - Prohibited securities.

Code of Federal Regulations, 2014 CFR

2014-01-01

... licenses for facilities which generate electric energy by means of a nuclear reactor; (2) State or local... reactor or a low-level waste facility; (3) Entities manufacturing or selling nuclear power or test reactors; (4) Architectural-engineering companies providing services relating to a nuclear power reactor...

5 CFR 5801.102 - Prohibited securities.

Code of Federal Regulations, 2010 CFR

2010-01-01

... licenses for facilities which generate electric energy by means of a nuclear reactor; (2) State or local... reactor or a low-level waste facility; (3) Entities manufacturing or selling nuclear power or test reactors; (4) Architectural-engineering companies providing services relating to a nuclear power reactor...

Reactor performance and microbial community dynamics during anaerobic co-digestion of municipal wastewater sludge with restaurant grease waste at steady state and overloading stages.

PubMed

Razaviarani, Vahid; Buchanan, Ian D

2014-11-01

Linkage between reactor performance and microbial community dynamics was investigated during mesophilic anaerobic co-digestion of restaurant grease waste (GTW) with municipal wastewater sludge (MWS) using 10L completely mixed reactors and a 20day SRT. Test reactors received a mixture of GTW and MWS while control reactors received only MWS. Addition of GTW to the test reactors enhanced the biogas production and methane yield by up to 65% and 120%, respectively. Pyrosequencing revealed that Methanosaeta and Methanomicrobium were the dominant acetoclastic and hydrogenotrophic methanogen genera, respectively, during stable reactor operation. The number of Methanosarcina and Methanomicrobium sequences increased and that of Methanosaeta declined when the proportion of GTW in the feed was increased to cause an overload condition. Under this overload condition, the pH, alkalinity and methane production decreased and VFA concentrations increased dramatically. Candidatus cloacamonas, affiliated within phylum Spirochaetes, were the dominant bacterial genus at all reactor loadings. Copyright © 2014 Elsevier Ltd. All rights reserved.

ETRCF, TRA654, INTERIOR. CAMERA IS ON MAIN FLOOR. NOTE CRANE ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR-CF, TRA-654, INTERIOR. CAMERA IS ON MAIN FLOOR. NOTE CRANE HOOKS. ELECTRICAL EQUIPMENT IS PART OF PAST EXPERIMENT. DOOR AT LEFT EDGE OF VIEW LEADS TO REACTOR SERVICE BUILDING, TRA-635. INL NEGATIVE NO. HD24-1-2. Mike Crane, Photographer, ca. 2003 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Parametric study of natural circulation flow in molten salt fuel in molten salt reactor

NASA Astrophysics Data System (ADS)

Pauzi, Anas Muhamad; Cioncolini, Andrea; Iacovides, Hector

2015-04-01

The Molten Salt Reactor (MSR) is one of the most promising system proposed by Generation IV Forum (GIF) for future nuclear reactor systems. Advantages of the MSR are significantly larger compared to other reactor system, and is mainly achieved from its liquid nature of fuel and coolant. Further improvement to this system, which is a natural circulating molten fuel salt inside its tube in the reactor core is proposed, to achieve advantages of reducing and simplifying the MSR design proposed by GIF. Thermal hydraulic analysis on the proposed system was completed using a commercial computation fluid dynamics (CFD) software called FLUENT by ANSYS Inc. An understanding on theory behind this unique natural circulation flow inside the tube caused by fission heat generated in molten fuel salt and tube cooling was briefly introduced. Currently, no commercial CFD software could perfectly simulate natural circulation flow, hence, modeling this flow problem in FLUENT is introduced and analyzed to obtain best simulation results. Results obtained demonstrate the existence of periodical transient nature of flow problem, hence improvements in tube design is proposed based on the analysis on temperature and velocity profile. Results show that the proposed system could operate at up to 750MW core power, given that turbulence are enhanced throughout flow region, and precise molten fuel salt physical properties could be defined. At the request of the authors and the Proceedings Editor the name of the co-author Andrea Cioncolini was corrected from Andrea Coincolini. The same name correction was made in the Acknowledgement section on page 030004-10 and in reference number 4. The updated article was published on 11 May 2015.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cappiello, M.; Hobbins, R.; Penny, K.

As part of the Department of Energy Advanced Fuel Cycle program, a series of fuels development irradiation tests have been performed in the Advanced Test Reactor (ATR) at the Idaho National Laboratory. These tests are providing excellent data for advanced fuels development. The program is focused on the transmutation of higher actinides which best can be accomplished in a sodium-cooled fast reactor. Because a fast test reactor is no longer available in the US, a special test vehicle is used to achieve near-prototypic fast reactor conditions (neutron spectra and temperature) for use in ATR (a water-cooled thermal reactor). As partmore » of the testing program, there were many successful tests of advanced fuels including metals and ceramics. Recently however, there have been three experimental campaigns using metal fuels that experienced failure during irradiation. At the request of the program, an independent review committee was convened to review the post-test analyses performed by the fuels development team, to assess the conclusions of the team for the cause of the failures, to assess the adequacy and completeness of the analyses, to identify issues that were missed, and to make recommendations for improvements in the design and operation of future tests. Although there is some difference of opinion, the review committee largely agreed with the conclusions of the fuel development team regarding the cause of the failures. For the most part, the analyses that support the conclusions are sufficient.« less

Naval Reactors Prime Contractor Team (NRPCT) Experiences and Considerations With Irradiation Test Performance in an International Environment

DOE Office of Scientific and Technical Information (OSTI.GOV)

MH Lane

2006-02-15

This letter forwards a compilation of knowledge gained regarding international interactions and issues associated with Project Prometheus. The following topics are discussed herein: (1) Assessment of international fast reactor capability and availability; (2) Japanese fast reactor (JOYO) contracting strategy; (3) NRPCT/Program Office international contract follow; (4) Completion of the Japan Atomic Energy Agency (JAEA)/Pacific Northwest National Laboratory (PNNL) contract for manufacture of reactor test components; (5) US/Japanese Departmental interactions and required Treaties and Agreements; and (6) Non-technical details--interactions and considerations.

In-reactor oxidation of zircaloy-4 under low water vapor pressures

NASA Astrophysics Data System (ADS)

Luscher, Walter G.; Senor, David J.; Clayton, Kevin K.; Longhurst, Glen R.

2015-01-01

Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330 and 370 °C). Data from these tests will be used to support the fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr-4 over the specified range of test conditions. Comparisons between in- and ex-reactor test results were performed to evaluate the influence of irradiation.

High yields of hydrogen production from methanol steam reforming with a cross-U type reactor

PubMed Central

Zhang, Shubin; Chen, Junyu; Zhang, Xuelin; Liu, Xiaowei

2017-01-01

This paper presents a numerical and experimental study on the performance of a methanol steam reformer integrated with a hydrogen/air combustion reactor for hydrogen production. A CFD-based 3D model with mass and momentum transport and temperature characteristics is established. The simulation results show that better performance is achieved in the cross-U type reactor compared to either a tubular reactor or a parallel-U type reactor because of more effective heat transfer characteristics. Furthermore, Cu-based micro reformers of both cross-U and parallel-U type reactors are designed, fabricated and tested for experimental validation. Under the same condition for reforming and combustion, the results demonstrate that higher methanol conversion is achievable in cross-U type reactor. However, it is also found in cross-U type reactor that methanol reforming selectivity is the lowest due to the decreased water gas shift reaction under high temperature, thereby carbon monoxide concentration is increased. Furthermore, the reformed gas generated from the reactors is fed into a high temperature proton exchange membrane fuel cell (PEMFC). In the test of discharging for 4 h, the fuel cell fed by cross-U type reactor exhibits the most stable performance. PMID:29121067

High yields of hydrogen production from methanol steam reforming with a cross-U type reactor.

PubMed

Zhang, Shubin; Zhang, Yufeng; Chen, Junyu; Zhang, Xuelin; Liu, Xiaowei

2017-01-01

This paper presents a numerical and experimental study on the performance of a methanol steam reformer integrated with a hydrogen/air combustion reactor for hydrogen production. A CFD-based 3D model with mass and momentum transport and temperature characteristics is established. The simulation results show that better performance is achieved in the cross-U type reactor compared to either a tubular reactor or a parallel-U type reactor because of more effective heat transfer characteristics. Furthermore, Cu-based micro reformers of both cross-U and parallel-U type reactors are designed, fabricated and tested for experimental validation. Under the same condition for reforming and combustion, the results demonstrate that higher methanol conversion is achievable in cross-U type reactor. However, it is also found in cross-U type reactor that methanol reforming selectivity is the lowest due to the decreased water gas shift reaction under high temperature, thereby carbon monoxide concentration is increased. Furthermore, the reformed gas generated from the reactors is fed into a high temperature proton exchange membrane fuel cell (PEMFC). In the test of discharging for 4 h, the fuel cell fed by cross-U type reactor exhibits the most stable performance.

A Review of Gas-Cooled Reactor Concepts for SDI Applications

DTIC Science & Technology

1989-08-01

710 program .) Wire- Core Reactor (proposed by Rockwell). The wire- core reactor utilizes thin fuel wires woven between spacer wires to form an open...reactor is based on results of developmental studies of nuclear rocket propulsion systems. The reactor core is made up of annular fuel assemblies of...XE Addendum to Volume II. NERVA Fuel Development , Westinghouse Astronuclear Laboratory, TNR-230, July 15’ 1972. J I8- Rover Program Reactor Tests

Responses to diagnostic tests for bovine tuberculosis in dairy and non-dairy cattle naturally exposed to Mycobacterium bovis in Great Britain.

PubMed

Downs, S H; Broughan, J M; Goodchild, A V; Upton, P A; Durr, P A

2016-10-01

Field surveillance of British cattle using the single intradermal comparative cervical tuberculin (SICCT) test shows a higher incidence rate of bovine tuberculosis (bTB) in dairy compared to beef herds, but a lower probability of post-mortem examination confirmed (PMC) Mycobacterium bovis infection in dairy herds. A cross-sectional study was conducted to compare animal level differences in bTB detection between dairy and non-dairy cattle in Great Britain. During the period from 2002 to 2005, 200 (41% dairy) reactors in the SICCT test (standard interpretation) were randomly selected, and 200 in-contact cattle (43% dairy) were purposively selected from bTB-infected herds. Interferon (IFN)-γ responses in blood to bovine and avian purified protein derivative (PPD), and early secretory antigen target 6 kDa and culture filtrate protein 10 (ESAT-6/CFP10), were measured. The post-mortem examination included gross pathological examination, mycobacterial culture and histopathology. The proportions of cattle positive to ESAT6/CFP10 were 26% (95% confidence interval, CI, 15-39%) in dairy reactors and 62% (95% CI 51-72%) in non-dairy reactors (P <0.001). PMC risk was 34% (95% CI 24-45%) in dairy reactors and 69% (95% CI 60-78%) in non-dairy reactors (P <0.001). The odds ratio for PMC risk in dairy reactors compared to non-dairy reactors, after controlling for bTB prevalence, herd size and SICCT test response, was 0.27 (95% CI 0.14-0.53; P <0.001). In surveillance data, adjusted animal level PMC risks were lower for dairy reactors than for beef reactors aged >2 years (P <0.001). Crown Copyright © 2016. Published by Elsevier Ltd. All rights reserved.

Fabrication and Testing of a Modular Micro-Pocket Fission Detector Instrumentation System for Test Nuclear Reactors

NASA Astrophysics Data System (ADS)

Reichenberger, Michael A.; Nichols, Daniel M.; Stevenson, Sarah R.; Swope, Tanner M.; Hilger, Caden W.; Roberts, Jeremy A.; Unruh, Troy C.; McGregor, Douglas S.

2018-01-01

Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Measurement of the neutron-flux distribution within the reactor core provides a more complete understanding of the operating conditions in the reactor than typical ex-core sensors. Micro-Pocket Fission Detectors have been developed and tested previously but have been limited to single-node operation and have utilized highly specialized designs. The development of a widely deployable, multi-node Micro-Pocket Fission Detector assembly will enhance nuclear research capabilities. A modular, four-node Micro-Pocket Fission Detector array was designed, fabricated, and tested at Kansas State University. The array was constructed from materials that do not significantly perturb the neutron flux in the reactor core. All four sensor nodes were equally spaced axially in the array to span the fuel-region of the reactor core. The array was filled with neon gas, serving as an ionization medium in the small cavities of the Micro-Pocket Fission Detectors. The modular design of the instrument facilitates the testing and deployment of numerous sensor arrays. The unified design drastically improved device ruggedness and simplified construction from previous designs. Five 8-mm penetrations in the upper grid plate of the Kansas State University TRIGA Mk. II research nuclear reactor were utilized to deploy the array between fuel elements in the core. The Micro-Pocket Fission Detector array was coupled to an electronic support system which has been specially developed to support pulse-mode operation. The Micro-Pocket Fission Detector array composed of four sensors was used to monitor local neutron flux at a constant reactor power of 100 kWth at different axial locations simultaneously. The array was positioned at five different radial locations within the core to emulate the deployment of multiple arrays and develop a 2-dimensional measurement of neutron flux in the reactor core.

Testing of a Transport Cask for Research Reactor Spent Fuel - 13003

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mourao, Rogerio P.; Leite da Silva, Luiz; Miranda, Carlos A.

2013-07-01

Since the beginning of the last decade three Latin American countries that operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the TRIGA and MTR reactors operated in the region. A main drive in this initiative, sponsored by the International Atomic Energy Agency, is the fact that no definite solution regarding the back end of the research reactor fuel cycle has been taken by any of the participating country. However, any long-term solution - either disposition in a repository or storage away frommore » reactor - will involve at some stage the transportation of the spent fuel through public roads. Therefore, a licensed cask that provides adequate shielding, assurance of subcriticality, and conformance to internationally accepted safety, security and safeguards regimes is considered a strategic part of any future solution to be adopted at a regional level. As a step in this direction, a packaging for the transport of irradiated fuel for MTR and TRIGA research reactors was designed by the tri-national team and a half-scale model equipped with the MTR version of the internal basket was constructed in Argentina and Brazil and tested in Brazil. Three test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. After failing the tests in the first two test series, the specimen successfully underwent the last test sequence. A second specimen, incorporating the structural improvements in view of the previous tests results, will be tested in the near future. Numerical simulations of the free drop and thermal tests are being carried out in parallel, in order to validate the computational modeling that is going to be used as a support for the package certification. (authors)« less

Multi-Purpose Thermal Hydraulic Loop: Advanced Reactor Technology Integral System Test (ARTIST) Facility for Support of Advanced Reactor Technologies

DOE Office of Scientific and Technical Information (OSTI.GOV)

James E. O'Brien; Piyush Sabharwall; SuJong Yoon

2001-11-01

Effective and robust high temperature heat transfer systems are fundamental to the successful deployment of advanced reactors for both power generation and non-electric applications. Plant designs often include an intermediate heat transfer loop (IHTL) with heat exchangers at either end to deliver thermal energy to the application while providing isolation of the primary reactor system. In order to address technical feasibility concerns and challenges a new high-temperature multi-fluid, multi-loop test facility “Advanced Reactor Technology Integral System Test facility” (ARTIST) is under development at the Idaho National Laboratory. The facility will include three flow loops: high-temperature helium, molten salt, and steam/water.more » Details of some of the design aspects and challenges of this facility, which is currently in the conceptual design phase, are discussed« less

78 FR 48501 - Agency Information Collection Activities: Proposed Collection; Comment Request

Federal Register 2010, 2011, 2012, 2013, 2014

2013-08-08

... storage installations, decommissioned power reactors, power reactors under construction, research and test reactors, agreement states, non-agreement states, as well as departments of health, medical centers, steel...

MTR,TRA603. EXPERIMENTERS' SPACE ALLOCATIONS IN BASEMENT AS OF 1963. SHIELDED ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

MTR,TRA-603. EXPERIMENTERS' SPACE ALLOCATIONS IN BASEMENT AS OF 1963. SHIELDED CUBICLES WERE IDENTIFIED BY SPONSORING LABORATORY AND ITS TEST HOLE NUMBER IN THE REACTOR, IE, "KAPL HB-1" SIGNIFIED KNOLLS ATOMIC POWER LABORATORY, HORIZONTAL BEAM NO. 1. "WAPD" WAS WESTINGHOUSE ATOMIC POWER DIVISION. CATCH TANKS AND SAMPLE STATIONS FOR TEST LOOPS WERE ASSOCIATED WITH THESE CUBICLES. NOTE DESKS, STORAGE CABINETS, SWITCH GEAR, INSTRUMENT PANELS. PHILLIPS PETROLEUM COMPANY MTR-E-5205, 4/1963. INL INDEX NO. 531-0603-00-706-009757, REV. 5. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Flow tests of a single fuel element coolant channel for a compact fast reactor for space power

NASA Technical Reports Server (NTRS)

Springborn, R. H.

1971-01-01

Water flow tests were conducted on a single-fuel-element cooling channel for a nuclear concept to be used for space power. The tests established a method for measuring coolant flow rate which is applicable to water flow testing of a complete mockup of the reference reactor. The inlet plenum-to-outlet plenum pressure drop, which approximates the overall core pressure drop, was measured and correlated with flow rate. This information can be used for reactor coolant flow and heat transfer calculations. An analytical study of the flow characteristics was also conducted.

Thermal exploitation of wastes with lignite for energy production.

PubMed

Grammelis, Panagiotis; Kakaras, Emmanuel; Skodras, George

2003-11-01

The thermal exploitation of wastewood with Greek lignite was investigated by performing tests in a laboratory-scale fluidized bed reactor, a 1-MW(th) semi-industrial circulating fluidized bed combustor, and an industrial boiler. Blends of natural wood, demolition wood, railroad sleepers, medium-density fiberboard residues, and power poles with lignite were used, and the co-combustion efficiency and the effect of wastewood addition on the emitted pollutants were investigated. Carbon monoxide, sulfur dioxide, and oxides of nitrogen emissions were continuously monitored, and, during the industrial-scale tests, the toxic emissions (polychlorinated dibenzodioxins and dibenzofurans and heavy metals) were determined. Ash samples were analyzed for heavy metals in an inductively coupled plasma-atomic emission spectroscopy spectrophotometer. Problems were observed during the preparation of wastewood, because species embedded with different compounds, such as railway sleepers and demolition wood, were not easily treated. All wastewood blends were proven good fuels; co-combustion proceeded smoothly and homogeneous temperature and pressure profiles were obtained. Although some fluctuations were observed, low emissions of gaseous pollutants were obtained for all fuel blends. The metal element emissions (in the flue gases and the solid residues) were lower than the legislative limits. Therefore, wastewood co-combustion with lignite can be realized, provided that the fuel handling and preparation can be practically performed in large-scale installations.

Testing for a Sterile Neutrino in Computer Models of the RNPS Short Baseline Nuclear Reactor Experiments.

NASA Astrophysics Data System (ADS)

Hunter, Rachel; Ernst, David; Vastola, John; Austin, Noah

In the 1980's and 90's a series of experiments were conducted to search for evidence of neutrino oscillations. Data was collected on five of the six independent fundamental parameters relating to oscillation rates. The data was then used to produce an exclusion region plot for values of the parameters. However, it was discovered that the experiments were not analyzed correctly and there are large gaps between theoretical and experimental data. A fourth type of neutrino could be to blame for these gaps. The goal of this research project is to find evidence for or against a fourth type of neutrino by a reanalysis of the old experiments. This part of the project attempts to reproduce the exclusion region plots for data taken at Rovno Nuclear Power Station in order to validate a model of the original analysis. Thus far the reproduction of their exclusion region is close, but not a complete success. Further work on the coding program will need to be completed in order to proceed with the next step in the reanalysis procedure. National Science Foundation Grant #1263045.

Oxidation kinetics of molten copper sulfide

NASA Astrophysics Data System (ADS)

Alyaser, A. H.; Brimacombe, J. K.

1995-02-01

The oxidation kinetics of molten Cu2S baths, during top lancing with oxygen/nitrogen (argon) mixtures, have been investigated as a function of oxygen partial pressure (0.2 to 0.78), bath temperature (1200 °C to 1300 °C), gas flow rate (1 to 4 L/min), and bath mixing. Surface-tension-driven flows (the Marangoni effect) were observed both visually and photographically. Thus, the oxidation of molten Cu2S was found to progress in two distinct stages, the kinetics of which are limited by the mass transfer of oxygen in the gas phase to the melt surface. During the primary stage, the melt is partially desulfurized while oxygen dissolves in the liquid sulfide. Upon saturation of the melt with oxygen, the secondary stage commences in which surface and bath reactions proceed to generate copper and SO2 electrochemically. A mathematical model of the reaction kinetics has been formulated and tested against the measurements. The results of this study shed light on the process kinetics of the copper blow in a Peirce-Smith converter or Mitsubishi reactor.

Enhanced biological nutrient removal in sequencing batch reactors operated as static/oxic/anoxic (SOA) process.

PubMed

Xu, Dechao; Chen, Hongbo; Li, Xiaoming; Yang, Qi; Zeng, Tianjing; Luo, Kun; Zeng, Guangming

2013-09-01

An innovative static/oxic/anoxic (SOA) activated sludge process characterized by static phase as a substitute for conventional anaerobic stage was developed to enhance biological nutrient removal (BNR) with influent ammonia of 20 and 40 mg/L in R1 and R2 reactors, respectively. The results demonstrated that static phase could function as conventional anaerobic stage. In R1 lower influent ammonia concentration facilitated more polyphosphate accumulating organisms (PAOs) growth, but secondary phosphorus release occurred due to NOx(-) depletion during post-anoxic period. In R2, however, denitrifying phosphorus removal proceeded with sufficient NOx(-). Both R1 and R2 saw simultaneous nitrification-denitrification. Glycogen was utilized to drive post-denitrification with denitrification rates in excess of typical endogenous decay rates. The anoxic stirring duration could be shortened from 3 to 1.5h to avoid secondary phosphorus release in R1 and little adverse impact was found on nutrients removal in R2. Copyright © 2013 Elsevier Ltd. All rights reserved.

Effect of COD/SO4(2-) ratio on anaerobic treatment of landfill leachate during the start-up period.

PubMed

Yilmaz, Tuba; Erdirencelebi, Dilek; Berktay, Ali

2012-01-01

This study investigates the performance of an anaerobic baffled reactor (ABR) during the start-up period of raw young landfill leachate treatment at two chemical oxygen demand (COD) to SO4(2-) ratios of 20 and 4. The reactor was operated at ambient temperature and low organic loading rates (0.52, 0.76 and 1.05 kg COD/m3 per day). During the study, sulfate-reducing bacteria (SRB) activity increased at the lower ratio of COD/SO4(2-) producing higher levels of sulfide and alkalinity. The dissolved sulfide concentration reached an inhibitory level above 250 mg/L, which caused a sharp reduction in the total COD removal efficiency from 77-80% to 32%. Total volatile fatty acid (TVFA) production proceeded at a constant level despite increased organic loading. As the effluent total and organic COD concentrations increased, the inhibitory effect of the inborn sulfide was correlated to the limitation experienced in the hydrolysis/acidogenesis stages, and thus VFA production and organic matter removal.

Enhanced In-Pile Instrumentation at the Advanced Test Reactor

NASA Astrophysics Data System (ADS)

Rempe, Joy L.; Knudson, Darrell L.; Daw, Joshua E.; Unruh, Troy; Chase, Benjamin M.; Palmer, Joe; Condie, Keith G.; Davis, Kurt L.

2012-08-01

Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory. To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility in 2007 to support the development and deployment of enhanced in-pile sensors. This paper provides an update on this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and real-time flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted.

ETR HEAT EXCHANGER BUILDING, TRA644. FLOOR PLAN AND SECTIONS. PUMP ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR HEAT EXCHANGER BUILDING, TRA-644. FLOOR PLAN AND SECTIONS. PUMP CUBICLES WITH PUMP MOTORS OUTSIDE CUBICLES. HEAT EXCHANGER EQUIPMENT. COOLANT PIPE TUNNEL ENTERS FROM REACTOR BUILDING. KAISER ETR-5582-MTR-644-A-3, 2/1956. INL INDEX NO. 532-0644-00-486-101294, REV. 6. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

High-Temperature Gas-Cooled Test Reactor Point Design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sterbentz, James William; Bayless, Paul David; Nelson, Lee Orville

2016-04-01

A point design has been developed for a 200 MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technological readiness level, licensing approach and costs.

ETR, TRA642. NORTHSOUTH SECTION, LOOKING WEST. STEELFRAME ROOF, CRANE RAIL, ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR, TRA-642. NORTH-SOUTH SECTION, LOOKING WEST. STEEL-FRAME ROOF, CRANE RAIL, AND CRANES. COOLANT PIPE TUNNEL LEADING TO REACTOR FROM EAST. (THIS WAS A PRELIMINARY CONCEPT DRAWING.) KAISER ETR-5528-MTR-642-A-4, 11/1955. INL INDEX NO. 532-0642-00-486-100912, REV. 1. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

10 CFR 100.10 - Factors to be considered when evaluating sites.

Code of Federal Regulations, 2011 CFR

2011-01-01

... Section 100.10 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) REACTOR SITE CRITERIA Evaluation Factors for Stationary Power Reactor Site Applications Before January 10, 1997 and for Testing Reactors § 100... include those relating both to the proposed reactor design and the characteristics peculiar to the site...

10 CFR 100.10 - Factors to be considered when evaluating sites.

Code of Federal Regulations, 2010 CFR

2010-01-01

... Section 100.10 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) REACTOR SITE CRITERIA Evaluation Factors for Stationary Power Reactor Site Applications Before January 10, 1997 and for Testing Reactors § 100... include those relating both to the proposed reactor design and the characteristics peculiar to the site...

REACTOR FUEL ELEMENTS TESTING CONTAINER

DOEpatents

Whitham, G.K.; Smith, R.R.

1963-01-15

This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

Progress towards developing neutron tolerant magnetostrictive and piezoelectric transducers

NASA Astrophysics Data System (ADS)

Reinhardt, Brian; Tittmann, Bernhard; Rempe, Joy; Daw, Joshua; Kohse, Gordon; Carpenter, David; Ames, Michael; Ostrovsky, Yakov; Ramuhalli, Pradeep; Montgomery, Robert; Chien, Hualte; Wernsman, Bernard

2015-03-01

Current generation light water reactors (LWRs), sodium cooled fast reactors (SFRs), small modular reactors (SMRs), and next generation nuclear plants (NGNPs) produce harsh environments in and near the reactor core that can severely tax material performance and limit component operational life. To address this issue, several Department of Energy Office of Nuclear Energy (DOE-NE) research programs are evaluating the long duration irradiation performance of fuel and structural materials used in existing and new reactors. In order to maximize the amount of information obtained from Material Testing Reactor (MTR) irradiations, DOE is also funding development of enhanced instrumentation that will be able to obtain in-situ, real-time data on key material characteristics and properties, with unprecedented accuracy and resolution. Such data are required to validate new multi-scale, multi-physics modeling tools under development as part of a science-based, engineering driven approach to reactor development. It is not feasible to obtain high resolution/microscale data with the current state of instrumentation technology. However, ultrasound-based sensors offer the ability to obtain such data if it is demonstrated that these sensors and their associated transducers are resistant to high neutron flux, high gamma radiation, and high temperature. To address this need, the Advanced Test Reactor National Scientific User Facility (ATR-NSUF) is funding an irradiation, led by PSU, at the Massachusetts Institute of Technology Research Reactor to test the survivability of ultrasound transducers. As part of this effort, PSU and collaborators have designed, fabricated, and provided piezoelectric and magnetostrictive transducers that are optimized to perform in harsh, high flux, environments. Four piezoelectric transducers were fabricated with either aluminum nitride, zinc oxide, or bismuth titanate as the active element that were coupled to either Kovar or aluminum waveguides and two magnetostrictive transducers were fabricated with Remendur or Galfenol as the active elements. Pulse-echo ultrasonic measurements of these transducers are made in-situ. This paper will present an overview of the test design including selection criteria for candidate materials and optimization of test assembly parameters, data obtained from both out-of-pile and in-pile testing at elevated temperatures, and an assessment based on initial data of the expected performance of ultrasonic devices in irradiation conditions.

49 CFR 40.385 - Who bears the burden of proof in a PIE proceeding?

Code of Federal Regulations, 2010 CFR

2010-10-01

... 49 Transportation 1 2010-10-01 2010-10-01 false Who bears the burden of proof in a PIE proceeding... TRANSPORTATION WORKPLACE DRUG AND ALCOHOL TESTING PROGRAMS Public Interest Exclusions § 40.385 Who bears the burden of proof in a PIE proceeding? (a) As the proponent of issuing a PIE, the initiating official bears...

Modeling of the Reactor Core Isolation Cooling Response to Beyond Design Basis Operations - Interim Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ross, Kyle; Cardoni, Jeffrey N.; Wilson, Chisom Shawn

2015-12-01

Efforts are being pursued to develop and qualify a system-level model of a reactor core isolation (RCIC) steam-turbine-driven pump. The model is being developed with the intent of employing it to inform the design of experimental configurations for full-scale RCIC testing. The model is expected to be especially valuable in sizing equipment needed in the testing. An additional intent is to use the model in understanding more fully how RCIC apparently managed to operate far removed from its design envelope in the Fukushima Daiichi Unit 2 accident. RCIC modeling is proceeding along two avenues that are expected to complement eachmore » other well. The first avenue is the continued development of the system-level RCIC model that will serve in simulating a full reactor system or full experimental configuration of which a RCIC system is part. The model reasonably represents a RCIC system today, especially given design operating conditions, but lacks specifics that are likely important in representing the off-design conditions a RCIC system might experience in an emergency situation such as a loss of all electrical power. A known specific lacking in the system model, for example, is the efficiency at which a flashing slug of water (as opposed to a concentrated jet of steam) could propel the rotating drive wheel of a RCIC turbine. To address this specific, the second avenue is being pursued wherein computational fluid dynamics (CFD) analyses of such a jet are being carried out. The results of the CFD analyses will thus complement and inform the system modeling. The system modeling will, in turn, complement the CFD analysis by providing the system information needed to impose appropriate boundary conditions on the CFD simulations. The system model will be used to inform the selection of configurations and equipment best suitable of supporting planned RCIC experimental testing. Preliminary investigations with the RCIC model indicate that liquid water ingestion by the turbine decreases the developed turbine torque; the RCIC speed then slows, and thus the pump flow rate to the RPV decreases. Subsequently, RPV water level decreases due to continued boiling and the liquid fraction flowing to the RCIC decreases, thereby accelerating the RCIC and refilling the RPV. The feedback cycle then repeats itself and/or reaches a quasi-steady equilibrium condition. In other words, the water carry-over is limited by cyclic RCIC performance degradation, and hence the system becomes self-regulating. The indications achieved to date with the system model are more qualitative than quantitative. The avenues being pursued to increase the fidelity of the model are expected to add quantitative realism. The end product will be generic in the sense that the RCIC model will be incorporable within the larger reactor coolant system model of any nuclear power plant or experimental configuration.« less

A unifying picture of gas-phase formation and growth of PAH (Polycyclic Aromatic Hydrocarbons), soot, diamond and graphite

NASA Technical Reports Server (NTRS)

Frenklach, Michael

1990-01-01

A variety of seemingly different carbon formation processes -- polycyclic aromatic hydrocarbons and diamond in the interstellar medium, soot in hydrocarbon flames, graphite and diamond in plasma-assisted-chemical vapor deposition reactors -- may all have closely related underlying chemical reaction mechanisms. Two distinct mechanisms for gas-phase carbon growth are discussed. At high temperatures it proceeds via the formation of carbon clusters. At lower temperatures it follows a polymerization-type kinetic sequence of chemical reactions of acetylene addition to a radical, and reactivation of the resultant species through H-abstraction by a hydrogen atom.

Proceedings of the Conference on the Environmental Chemistry of Hydrazine Fuels (3rd) Held in Panama City Beach, Florida on 15-17 September 1987

DTIC Science & Technology

1988-01-01

for hydrauine, MMH and UDMH are 4.78 x 10-6, 10.2 x 10Ś, and 3.19 x 10-6 aecŕ, respectively. Plots of the log(area) versus time were linear and...followed first-order kinetics except for hydrauine, for which a non- linear portion was observed in the first 6 to 8 hours. This portion of the decay...As a result, the prototype flow reactor can be represented to good approximation by a linear combination of point source solutions (Reference 19). The

Nuclear Science Symposium, 19th, and Nuclear Power Systems Symposium, 4th, Miami, Fla., December 6-8, 1972, Proceedings.

NASA Technical Reports Server (NTRS)

1973-01-01

Major topics covered include radiation monitoring instrumentation, nuclear circuits and systems, biomedical applications of nuclear radiation in diagnosis and therapy, plasma research for fusion power, reactor control and instrumentation, nuclear power standards, and applications of digital computers in nuclear power plants. Systems and devices for space applications are described, including the Apollo alpha spectrometer, a position sensitive detection system for UV and X-ray photons, a 4500-volt electron multiplier bias supply for satellite use, spark chamber systems, proportional counters, and other devices. Individual items are announced in this issue.

Review of Nuclear Thermal Propulsion Ground Test Options

NASA Technical Reports Server (NTRS)

Coote, David J.; Power, Kevin P.; Gerrish, Harold P.; Doughty, Glen

2015-01-01

High efficiency rocket propulsion systems are essential for humanity to venture beyond the moon. Nuclear Thermal Propulsion (NTP) is a promising alternative to conventional chemical rockets with relatively high thrust and twice the efficiency of highest performing chemical propellant engines. NTP utilizes the coolant of a nuclear reactor to produce propulsive thrust. An NTP engine produces thrust by flowing hydrogen through a nuclear reactor to cool the reactor, heating the hydrogen and expelling it through a rocket nozzle. The hot gaseous hydrogen is nominally expected to be free of radioactive byproducts from the nuclear reactor; however, it has the potential to be contaminated due to off-nominal engine reactor performance. NTP ground testing is more difficult than chemical engine testing since current environmental regulations do not allow/permit open air testing of NTP as was done in the 1960's and 1970's for the Rover/NERVA program. A new and innovative approach to rocket engine ground test is required to mitigate the unique health and safety risks associated with the potential entrainment of radioactive waste from the NTP engine reactor core into the engine exhaust. Several studies have been conducted since the ROVER/NERVA program in the 1970's investigating NTP engine ground test options to understand the technical feasibility, identify technical challenges and associated risks and provide rough order of magnitude cost estimates for facility development and test operations. The options can be divided into two distinct schemes; (1) real-time filtering of the engine exhaust and its release to the environment or (2) capture and storage of engine exhaust for subsequent processing.

Impact of Substratum Surface on Microbial Community Structure and Treatment Performance in Biological Aerated Filters

PubMed Central

Kim, Lavane; Pagaling, Eulyn; Zuo, Yi Y.

2014-01-01

The impact of substratum surface property change on biofilm community structure was investigated using laboratory biological aerated filter (BAF) reactors and molecular microbial community analysis. Two substratum surfaces that differed in surface properties were created via surface coating and used to develop biofilms in test (modified surface) and control (original surface) BAF reactors. Microbial community analysis by 16S rRNA gene-based PCR-denaturing gradient gel electrophoresis (DGGE) showed that the surface property change consistently resulted in distinct profiles of microbial populations during replicate reactor start-ups. Pyrosequencing of the bar-coded 16S rRNA gene amplicons surveyed more than 90% of the microbial diversity in the microbial communities and identified 72 unique bacterial species within 19 bacterial orders. Among the 19 orders of bacteria detected, Burkholderiales and Rhodocyclales of the Betaproteobacteria class were numerically dominant and accounted for 90.5 to 97.4% of the sequence reads, and their relative abundances in the test and control BAF reactors were different in consistent patterns during the two reactor start-ups. Three of the five dominant bacterial species also showed consistent relative abundance changes between the test and control BAF reactors. The different biofilm microbial communities led to different treatment efficiencies, with consistently higher total organic carbon (TOC) removal in the test reactor than in the control reactor. Further understanding of how surface properties affect biofilm microbial communities and functional performance would enable the rational design of new generations of substrata for the improvement of biofilm-based biological treatment processes. PMID:24141134

Design test request No. 1263 K Reactor graphite key and VSR channel sleeve test

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kempf, F.J.

1964-12-10

The objectives of this test were: (1) Determine the coefficient of friction between two adjacent layers of K Reactor graphite at room temperature. (2) Determine the average load required to cause failure of an unirradiated K Reactor side reflector bar, when subjected to tensile loading applied through the reflector keys. (3) Determine the average load at failure and the average deflection at failure of a single VSR channel key when loaded in keyways with clearances equal to those used in original stack construction. (4) Determine the average load and deflection required to break the four K Reactor VSR keys whenmore » loaded simultaneously in both `3-layer` and `7-layer` mockups. Also determine the mode of key failure; i.e., shear, flexure or combined compression and bending. Following these key rupture tests, determine the strength and deflection characteristics of the proposed K Reactor VSR channel sleeve when loaded in a manner identical to that used to fracture the keys. (5) Determine the average load and deflection at failure of both the proposed K Reactor VSR channel sleeves and the proposed C Reactor sleeves when subjected to crushing loads. (6) Determine the extent of damage to the proposed K Reactor VSR channel sleeve when subjected to the following vertical rod loading conditions. (a) Full rod drop in a channel mockup which has been misaligned 2 1/2 inches. (b) Full rod drop in a channel which has been misaligned an amount equal to the maximum flexibility of a `universal` VSR.« less

U.S. Nuclear Cooperation with India: Issues for Congress

DTIC Science & Technology

2008-11-03

separation list: ! 8 indigenous Indian power reactors ! Fast Breeder test Reactor (FTBR) and Prototype Fast Breeder Reactors (PFBR) under construction...facilities like reprocessing and enrichment plants and breeder reactors could be viewed as providing a significant nonproliferation benefit because the... breeder reactors would support the 2002 U.S. National Strategy to Combat Weapons of Mass Destruction, in which the United States pledged to “continue to

U.S. Nuclear Cooperation with India: Issues for Congress

DTIC Science & Technology

2008-10-02

8 indigenous Indian power reactors ! Fast Breeder test Reactor (FTBR) and Prototype Fast Breeder Reactors (PFBR) under construction ! Enrichment... breeder reactors could be viewed as providing a significant nonproliferation benefit because the materials produced by these plants are a few steps closer...to potential use in a bomb. In addition, safeguards on enrichment, reprocessing plants, and breeder reactors would support the 2002 U.S. National

Gas phase oxidation downstream of a catalytic combustor

NASA Technical Reports Server (NTRS)

Tien, J. S.; Anderson, D. N.

1979-01-01

Effect of the length available for gas-phase reactions downstream of the catalytic reactor on the emission of CO and unburned hydrocarbons was investigated. A premixed, prevaporized propane/air feed to a 12/cm/diameter catalytic/reactor test section was used. The catalytic reactor was made of four 2.5 cm long monolithic catalyst elements. Four water cooled gas sampling probes were located at positions between 0 and 22 cm downstream of the catalytic reactor. Measurements of unburned hydrocarbon, CO, and CO2 were made. Tests were performed with an inlet air temperature of 800 K, a reference velocity of 10 m/s, pressures of 3 and 600,000 Pa, and fuel air equivalence ratios of 0.14 to 0.24. For very lean mixtures, hydrocarbon emissions were high and CO continued to be formed downstream of the catalytic reactor. At the highest equivalence ratios tested, hydrocarbon levels were much lower and CO was oxidized to CO2 in the gas phase downstream. To achieve acceptable emissions, a downstream region several times longer than the catalytic reactor could be required.

Comparison of heavy metal toxicity in continuous flow and batch reactors

NASA Astrophysics Data System (ADS)

Sengor, S. S.; Gikas, P.; Moberly, J. G.; Peyton, B. M.; Ginn, T. R.

2009-12-01

The presence of heavy metals may significantly affect microbial growth. In many cases, small amounts of particular heavy metals may stimulate microbial growth; however, larger quantities may result in microbial growth reduction. Environmental parameters, such as growth pattern may alter the critical heavy metal concentration, above which microbial growth stimulation turns to growth inhibition. Thus, it is important to quantify the effects of heavy metals on microbial activity for understanding natural or manmade biological reactors, either in situ or ex situ. Here we compare the toxicity of Zn and Cu on Arthrobacter sp., a heavy metal tolerant microorganism, under continuous flow versus batch reactor operations. Batch and continuous growth tests of Arthrobacter sp. were carried out at various individual and combined concentrations of Zn and Cu. Biomass concentration (OD) was measured for both the batch and continuous reactors, whereas ATP, oxygen uptake rates and substrate concentrations were additionally measured for the continuous system. Results indicated that Cu was more toxic than Zn under all conditions for both systems. In batch reactors, all tested Zn concentrations up to 150 uM showed a stimulatory effect on microbial growth. However, in the case of mixed Zn and Cu exposures, the presence of Zn either eliminated (at the 50 uM level both Zn and Cu) or reduced by ~25% (at the 100 and 150 uM levels both Zn and Cu) the Cu-induced inhibition. In the continuous system, only one test involved combined Cu (40uM) and Zn (125uM) and this test showed similar results to the 40uM Cu continuous test, i.e., no reduction in inhibition. The specific ATP concentration, i.e., ATP/OD, results for the continuous reactor showed an apparent recovery for both Cu-treated populations, although neither the OD nor glucose data showed any recovery. This may reflect that the individual microorganisms that survived after the addition of heavy metals, kept maintaining the usual ATP levels, as before metal addition. The last may imply a short of adaptation by some microorganisms to the presence of heavy metals. Overall, the batch reactor tests underestimated significantly the heavy metal inhibition, as compared to the continuous flow reactors. Therefore, the results of batch reactor tests should be used with some caution when heavy metal inhibition is to be interpreted for continuous flow natural environmental systems, such as rivers or wetlands.

Design and Laboratory Evaluation of Future Elongation and Diameter Measurements at the Advanced Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

K. L. Davis; D. L. Knudson; J. L. Rempe

New materials are being considered for fuel, cladding, and structures in next generation and existing nuclear reactors. Such materials can undergo significant dimensional and physical changes during high temperature irradiations. In order to accurately predict these changes, real-time data must be obtained under prototypic irradiation conditions for model development and validation. To provide such data, researchers at the Idaho National Laboratory (INL) High Temperature Test Laboratory (HTTL) are developing several instrumented test rigs to obtain data real-time from specimens irradiated in well-controlled pressurized water reactor (PWR) coolant conditions in the Advanced Test Reactor (ATR). This paper reports the status ofmore » INL efforts to develop and evaluate prototype test rigs that rely on Linear Variable Differential Transformers (LVDTs) in laboratory settings. Although similar LVDT-based test rigs have been deployed in lower flux Materials Testing Reactors (MTRs), this effort is unique because it relies on robust LVDTs that can withstand higher temperatures and higher fluxes than often found in other MTR irradiations. Specifically, the test rigs are designed for detecting changes in length and diameter of specimens irradiated in ATR PWR loops. Once implemented, these test rigs will provide ATR users with unique capabilities that are sorely needed to obtain measurements such as elongation caused by thermal expansion and/or creep loading and diameter changes associated with fuel and cladding swelling, pellet-clad interaction, and crud buildup.« less

Design and testing of the reactor-internal hydraulic control rod drive for the nuclear heating plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Batheja, P.; Meier, W.J.; Rau, P.J.

A hydraulically driven control rod is being developed at Kraftwerk Union for integration in the primary system of a small nuclear district heating reactor. An elaborate test program, under way for --3 yr, was initiated with a plexiglass rig to understand the basic principles. A design specification list was prepared, taking reactor boundary conditions and relevant German rules and regulations into account. Subsequently, an atmospheric loop for testing of components at 20 to 90/sup 0/C was erected. The objectives involved optimization of individual components such as a piston/cylinder drive unit, electromagnetic valves, and an ultrasonic position indication system as wellmore » as verification of computer codes. Based on the results obtained, full-scale components were designed and fabricated for a prototype test rig, which is currently in operation. Thus far, all atmospheric tests in this rig have been completed. Investigations under reactor temperature and pressure, followed by endurance tests, are under way. All tests to date have shown a reliable functioning of the hydraulic drive, including a novel ultrasonic position indication system.« less

A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

NASA Astrophysics Data System (ADS)

Reed, Mark; Parker, Ronald R.; Forget, Benoit

2012-06-01

This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritium allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more proliferation-resistant than that bred by conventional fast reactors. Furthermore, it can maintain constant total hybrid power output as burnup proceeds by varying the neutron source strength.

Evaluation of some candidate materials for automobile thermal reactors in engine-dynamometer screening tests

NASA Technical Reports Server (NTRS)

Oldrieve, R. E.

1971-01-01

Fourteen materials were evaluated in engine screening tests on full-size thermal reactors for automobile engine pollution control systems. Cyclic test-stand engine operation provided 2 hours at 1040 C and a 20-minute air-cool to 70 C each test cycle. Each reactor material was exposed to 83 cycles in 200 hours of engine testing. On the basis of resistance to oxidation and distortion, the best materials included two ferritic iron alloys (Ge 1541 and Armco 18S/R), several commercial oxidation-resistant coatings on AlSl 651 (19-9 DL), and possibly uncoated AISI 310. The best commercial coatings were Cr-Al, Ni-Cr, and a glass ceramic.

ETR, TRA642. CONSOLE FLOOR. CAMERA IS ON WEST SIDE OF ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR, TRA-642. CONSOLE FLOOR. CAMERA IS ON WEST SIDE OF FLOOR AND FACES NORTH. OUTER WALL OF STORAGE CANAL IS AT RIGHT. SHIELDING IS THICKER AT LOWER LEVEL, WHERE SPENT FUEL ELEMENTS WILL COOL AFTER REMOVAL FROM REACTOR. INL NEGATIVE NO. 56-1401. Jack L. Anderson, Photographer, 5/1/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR CRITICAL FACILITY, TRA654. SCIENTISTS STAND AT EDGE OF TANK ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR CRITICAL FACILITY, TRA-654. SCIENTISTS STAND AT EDGE OF TANK AND LIFT REMOVABLE BRIDGE ABOVE THE REACTOR. CONTROL RODS AND FUEL RODS ARE BELOW ENOUGH WATER TO SHIELD WORKERS ABOVE. NOTE CRANE RAILS ALONG WALLS, PUMICE BLOCK WALLS. INL NEGATIVE NO. 57-3690. R.G. Larsen, Photographer, 7/29/1957 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR CONTROL BUILDING, TRA647, INTERIOR. CONTROL ROOM, CONTEXTUAL VIEW. INSTRUMENT ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR CONTROL BUILDING, TRA-647, INTERIOR. CONTROL ROOM, CONTEXTUAL VIEW. INSTRUMENT PANELS AT REAR OF OPERATOR'S CONSOLE GAVE OPERATOR STATUS OF REACTOR PERFORMANCE, COOLANT-WATER CHARACTERISTICS AND OTHER INDICATORS. WINDOWS AT RIGHT LOOKED INTO ETR BUILDING FIRST FLOOR. CAMERA FACING EAST. INL NEGATIVE NO. HD42-6. Mike Crane, Photographer, 3/2004 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Electric-stepping-motor tests for a control-drum actuator of a nuclear reactor

NASA Technical Reports Server (NTRS)

Kieffer, A. W.

1972-01-01

Experimental tests were conducted on two stepping motors for application as reactor control-drum actuators. Various control-drum loads with frictional resistances ranging from approximately zero to 40 N-m and inertias ranging from zero to 0.424 kg-sq m were tested.

Results from Mechanical Testing of Silicon Carbide for Space Applications: Non-Destructive Evalution Samples and MISSE-6 Experiment Samples

DTIC Science & Technology

2010-06-07

the materials properties of silicon carbide plates”, S. Kenderian et al., 2009 SPIE Proceedings, vol. 7425 • Materials – 10” x 16” SiC plates...CONFERENCE PROCEEDING 3. DATES COVERED (From - To) 2008-2010 4. TITLE AND SUBTITLE Results from Mechanical Testing of Silicon Carbide for Space...for silicon carbide optical systems that covers material verification through system development. Recent laboratory results for testing of materials

Loss-of-Flow and Loss-of-Pressure Simulations of the BR2 Research Reactor with HEU and LEU Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Licht, J.; Bergeron, A.; Dionne, B.

2016-01-01

Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The reactor core of BR2 is located inside a pressure vessel that contains 79 channels in a hyperboloid configuration. The core configuration is highly variable as each channel can contain a fuel assembly, a control or regulating rod, an experimentalmore » device, or a beryllium or aluminum plug. Because of this variability, a representative core configuration, based on current reactor use, has been defined for the fuel conversion analyses. The code RELAP5/Mod 3.3 was used to perform the transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. The input model has been modernized relative to that historically used at BR2 taking into account the best modeling practices developed by Argonne National Laboratory (ANL) and BR2 engineers.« less

A novel plant protection strategy for transient reactors

NASA Astrophysics Data System (ADS)

Bhattacharyya, Samit K.; Lipinski, Walter C.; Hanan, Nelson A.

The present plant protection system (PPS) has been defined for use in the TREAT-upgrade (TU) reactor for controlled transient operation of reactor-fuel behavior testing under simulated reactor-accident conditions. A PPS with energy-dependent trip set points lowered worst-case clad temperatures by as much as 180 K, relative to the use of conventional fixed-level trip set points. The multilayered multilevel protection strategy represents the state-of-the-art in terrestrial transient reactor protection systems, and should be applicable to multi-MW space reactors.

Series-Bosch Technology for Oxygen Recovery During Lunar or Martian Surface Missions

NASA Technical Reports Server (NTRS)

Abney, Morgan B.; Mansell, J. Matthew; Rabenberg, Ellen; Stanley, Christine M.; Edmunson, Jennifer; Alleman, James E.; Chen, Kevin; Dumez, Sam

2014-01-01

Long-duration surface missions to the Moon or Mars will require life support systems that maximize resource recovery to minimize resupply from Earth. To address this need, NASA previously proposed a Series-Bosch (S-Bosch) oxygen recovery system, based on the Bosch process, which can theoretically recover 100% of the oxygen from metabolic carbon dioxide. Bosch processes have the added benefits of the potential to recover oxygen from atmospheric carbon dioxide and the use of regolith materials as catalysts, thereby eliminating the need for catalyst resupply from Earth. In 2012, NASA completed an initial design for an S-Bosch development test stand that incorporates two catalytic reactors in series including a Reverse Water-Gas Shift (RWGS) Reactor and a Carbon Formation Reactor (CFR). In 2013, fabrication of system components, with the exception of a CFR, and assembly of the test stand was initiated. Stand-alone testing of the RWGS reactor was completed to compare performance with design models. Continued testing of Lunar and Martian regolith simulants provided sufficient data to design a CFR intended to utilize these materials as catalysts. Finally, a study was conducted to explore the possibility of producing bricks from spent regolith catalysts. The results of initial demonstration testing of the RWGS reactor, results of continued catalyst performance testing of regolith simulants, and results of brick material properties testing are reported. Additionally, design considerations for a regolith-based CFR are discussed.

Series-Bosch Technology for Oxygen Recovery During Lunar or Martian Surface Missions

NASA Technical Reports Server (NTRS)

Abney, Morgan B.; Mansell, James M.; Stanley, Christine; Edmunson, Jennifer; Dumez, Samuel; Chen, Kevin; Alleman, James E.

2014-01-01

Long-duration surface missions to the Moon or Mars will require life support systems that maximize resource recovery to minimize resupply from Earth. To address this need, NASA previously proposed a Series-Bosch (S-Bosch) oxygen recovery system, based on the Bosch process, which can theoretically recover 100% of the oxygen from metabolic carbon dioxide. Bosch processes have the added benefits of the potential to recover oxygen from atmospheric carbon dioxide and the use of regolith materials as catalysts, thereby eliminating the need for catalyst resupply from Earth. In 2012, NASA completed an initial design for an S-Bosch development test stand that incorporates two catalytic reactors in series including a Reverse Water-Gas Shift (RWGS) Reactor and a Carbon Formation Reactor (CFR). In 2013, fabrication of system components, with the exception of a CFR, and assembly of the test stand was initiated. Stand-alone testing of the RWGS reactor was completed to compare performance with design models. Continued testing of Lunar and Martian regolith simulants provided sufficient data to design a CFR intended to utilize these materials as catalysts. Finally, a study was conducted to explore the possibility of producing bricks from spend regolith catalysts. The results of initial demonstration testing of the RWGS reactor, results of continued catalyst performance testing of regolith simulants, and results of brick material properties testing are reported. Additionally, design considerations for a regolith-based CFR are discussed.

ETR ELECTRICAL BUILDING, TRA648, INTERIOR. SWITCHGEAR. INL NEGATIVE NO. 563794. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR ELECTRICAL BUILDING, TRA-648, INTERIOR. SWITCHGEAR. INL NEGATIVE NO. 56-3794. Jack L. Anderson, Photographer, 11/26/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR ELECTRICAL BUILDING, TRA648. BATTERY ROOM. INL NEGATIVE NO. 563785. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR ELECTRICAL BUILDING, TRA-648. BATTERY ROOM. INL NEGATIVE NO. 56-3785. Jack L. Anderson, Photographer, 11/26/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

PROCESS WATER BUILDING, TRA605. INSIDE A FLASH EVAPORATOR. INL NEGATIVE ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PROCESS WATER BUILDING, TRA-605. INSIDE A FLASH EVAPORATOR. INL NEGATIVE NO. 3323. Unknown Photographer, 9/12/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Safety considerations in testing a fuel-rich aeropropulsion gas generator

NASA Technical Reports Server (NTRS)

Rollbuhler, R. James; Hulligan, David D.

1991-01-01

A catalyst containing reactor is being tested using a fuel-rich mixture of Jet A fuel and hot input air. The reactor product is a gaseous fuel that can be utilized in aeropropulsion gas turbine engines. Because the catalyst material is susceptible to damage from high temperature conditions, fuel-rich operating conditions are attained by introducing the fuel first into an inert gas stream in the reactor and then displacing the inert gas with reaction air. Once a desired fuel-to-air ratio is attained, only limited time is allowed for a catalyst induced reaction to occur; otherwise the inert gas is substituted for the air and the fuel flow is terminated. Because there presently is not a gas turbine combustor in which to burn the reactor product gas, the gas is combusted at the outlet of the test facility flare stack. This technique in operations has worked successfully in over 200 tests.

N.S. SAVANNAH, DRAFT OF FINAL SAFEGUARDS REPORT TEST, START-UP AND TRIALS, NEW YORK SHIPBUILDING CORPORATION

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1960-04-01

The N. S. Savannah program for testing, start-up, and initial operation of all reactor and propulsion components and systems is discussed. Definitions of test phases are given and various stages of the test program are outlined. A list of tests for the various reactor, propulsion, and other system components is included. (C.J.G.)

Micro-reactors for characterization of nanostructure-based sensors.

PubMed

Savu, R; Silveira, J V; Flacker, A; Vaz, A R; Joanni, E; Pinto, A C; Gobbi, A L; Santos, T E A; Rotondaro, A L P; Moshkalev, S A

2012-05-01

Fabrication and testing of micro-reactors for the characterization of nanosensors is presented in this work. The reactors have a small volume (100 μl) and are equipped with gas input/output channels. They were machined from a single piece of kovar in order to avoid leaks in the system due to additional welding. The contact pins were electrically insulated from the body of the reactor using a borosilicate sealing glass and the reactor was hermetically sealed using a lid and an elastomeric o-ring. One of the advantages of the reactor lies in its simple assembly and ease of use with any vacuum/gas system, allowing the connection of more than one device. Moreover, the lid can be modified in order to fit a window for in situ optical characterization. In order to prove its versatility, carbon nanotube-based sensors were tested using this micro-reactor. The devices were fabricated by depositing carbon nanotubes over 1 μm thick gold electrodes patterned onto Si/SiO(2) substrates. The sensors were tested using oxygen and nitrogen atmospheres, in the pressure range between 10(-5) and 10(-1) mbar. The small chamber volume allowed the measurement of fast sensor characteristic times, with the sensors showing good sensitivity towards gas and pressure as well as high reproducibility.

Micro-reactors for characterization of nanostructure-based sensors

NASA Astrophysics Data System (ADS)

Savu, R.; Silveira, J. V.; Flacker, A.; Vaz, A. R.; Joanni, E.; Pinto, A. C.; Gobbi, A. L.; Santos, T. E. A.; Rotondaro, A. L. P.; Moshkalev, S. A.

2012-05-01

Fabrication and testing of micro-reactors for the characterization of nanosensors is presented in this work. The reactors have a small volume (100 μl) and are equipped with gas input/output channels. They were machined from a single piece of kovar in order to avoid leaks in the system due to additional welding. The contact pins were electrically insulated from the body of the reactor using a borosilicate sealing glass and the reactor was hermetically sealed using a lid and an elastomeric o-ring. One of the advantages of the reactor lies in its simple assembly and ease of use with any vacuum/gas system, allowing the connection of more than one device. Moreover, the lid can be modified in order to fit a window for in situ optical characterization. In order to prove its versatility, carbon nanotube-based sensors were tested using this micro-reactor. The devices were fabricated by depositing carbon nanotubes over 1 μm thick gold electrodes patterned onto Si/SiO2 substrates. The sensors were tested using oxygen and nitrogen atmospheres, in the pressure range between 10-5 and 10-1 mbar. The small chamber volume allowed the measurement of fast sensor characteristic times, with the sensors showing good sensitivity towards gas and pressure as well as high reproducibility.

The advantages and disadvantages of using the TREAT reactor for nuclear laser experiments

NASA Astrophysics Data System (ADS)

Dickson, P. W.; Snyder, A. M.; Imel, G. R.; McConnell, R. J.

The Transient Reactor Test Facility (TREAT) is a large air-cooled test facility located at the Idaho National Engineering Laboratory. Two of the major design features of TREAT, its large size and its being an air-cooled reactor, provide clues to both its advantages and disadvantages for supporting nuclear laser experiments. Its large size, which is dictated by the dilute uranium/graphite fuel, permits accommodation of geometrically large experiments. However, TREAT's large size also results in relatively long transients so that the energy deposited in an experiment is large relative to the peak power available from the reactor. TREAT's air-cooling mode of operation allows its configuration to be changed fairly readily. Due to air cooling, the reactor cools down slowly, permitting only one full power transient a day, which can be a disadvantage in some experimental programs. The reactor is capable of both steady-state or transient operation.

Scaling Studies for Advanced High Temperature Reactor Concepts, Final Technical Report: October 2014—December 2017

DOE Office of Scientific and Technical Information (OSTI.GOV)

Woods, Brian; Gutowska, Izabela; Chiger, Howard

Computer simulations of nuclear reactor thermal-hydraulic phenomena are often used in the design and licensing of nuclear reactor systems. In order to assess the accuracy of these computer simulations, computer codes and methods are often validated against experimental data. This experimental data must be of sufficiently high quality in order to conduct a robust validation exercise. In addition, this experimental data is generally collected at experimental facilities that are of a smaller scale than the reactor systems that are being simulated due to cost considerations. Therefore, smaller scale test facilities must be designed and constructed in such a fashion tomore » ensure that the prototypical behavior of a particular nuclear reactor system is preserved. The work completed through this project has resulted in scaling analyses and conceptual design development for a test facility capable of collecting code validation data for the following high temperature gas reactor systems and events— 1. Passive natural circulation core cooling system, 2. pebble bed gas reactor concept, 3. General Atomics Energy Multiplier Module reactor, and 4. prismatic block design steam-water ingress event. In the event that code validation data for these systems or events is needed in the future, significant progress in the design of an appropriate integral-type test facility has already been completed as a result of this project. Where applicable, the next step would be to begin the detailed design development and material procurement. As part of this project applicable scaling analyses were completed and test facility design requirements developed. Conceptual designs were developed for the implementation of these design requirements at the Oregon State University (OSU) High Temperature Test Facility (HTTF). The original HTTF is based on a ¼-scale model of a high temperature gas reactor concept with the capability for both forced and natural circulation flow through a prismatic core with an electrical heat source. The peak core region temperature capability is 1400°C. As part of this project, an inventory of test facilities that could be used for these experimental programs was completed. Several of these facilities showed some promise, however, upon further investigation it became clear that only the OSU HTTF had the power and/or peak temperature limits that would allow for the experimental programs envisioned herein. Thus the conceptual design and feasibility study development focused on examining the feasibility of configuring the current HTTF to collect validation data for these experimental programs. In addition to the scaling analyses and conceptual design development, a test plan was developed for the envisioned modified test facility. This test plan included a discussion on an appropriate shakedown test program as well as the specific matrix tests. Finally, a feasibility study was completed to determine the cost and schedule considerations that would be important to any test program developed to investigate these designs and events.« less

Plum Brook Reactor Facility Control Room during Facility Startup

NASA Image and Video Library

1961-02-21

Operators test the National Aeronautics and Space Administration’s (NASA) Plum Brook Reactor Facility systems in the months leading up to its actual operation. The “Reactor On” signs are illuminated but the reactor core was not yet ready for chain reactions. Just a couple weeks after this photograph, Plum Brook Station held a media open house to unveil the 60-megawatt test reactor near Sandusky, Ohio. More than 60 members of the print media and radio and television news services met at the site to talk with community leaders and representatives from NASA and Atomic Energy Commission. The Plum Brook reactor went critical for the first time on the evening of June 14, 1961. It was not until April 1963 that the reactor reached its full potential of 60 megawatts. The reactor control room, located on the second floor of the facility, was run by licensed operators. The operators manually operated the shim rods which adjusted the chain reaction in the reactor core. The regulating rods could partially or completely shut down the reactor. The control room also housed remote area monitoring panels and other monitoring equipment that allowed operators to monitor radiation sensors located throughout the facility and to scram the reactor instantly if necessary. The color of the indicator lights corresponded with the elevation of the detectors in the various buildings. The reactor could also shut itself down automatically if the monitors detected any sudden irregularities.

Research Program of a Super Fast Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Oka, Yoshiaki; Ishiwatari, Yuki; Liu, Jie

2006-07-01

Research program of a supercritical-pressure light water cooled fast reactor (Super Fast Reactor) is funded by MEXT (Ministry of Education, Culture, Sports, Science and Technology) in December 2005 as one of the research programs of Japanese NERI (Nuclear Energy Research Initiative). It consists of three programs. (1) development of Super Fast Reactor concept; (2) thermal-hydraulic experiments; (3) material developments. The purpose of the concept development is to pursue the advantage of high power density of fast reactor over thermal reactors to achieve economic competitiveness of fast reactor for its deployment without waiting for exhausting uranium resources. Design goal is notmore » breeding, but maximizing reactor power by using plutonium from spent LWR fuel. MOX will be the fuel of the Super Fast Reactor. Thermal-hydraulic experiments will be conducted with HCFC22 (Hydro chlorofluorocarbons) heat transfer loop of Kyushu University and supercritical water loop at JAEA. Heat transfer data including effect of grid spacers will be taken. The critical flow and condensation of supercritical fluid will be studied. The materials research includes the development and testing of austenitic stainless steel cladding from the experience of PNC1520 for LMFBR. Material for thermal insulation will be tested. SCWR (Supercritical-Water Cooled Reactor) of GIF (Generation-4 International Forum) includes both thermal and fast reactors. The research of the Super Fast Reactor will enhance SCWR research and the data base. The research period will be until March 2010. (authors)« less

Design and Analysis of Embedded I&C for a Fully Submerged Magnetically Suspended Impeller Pump

DOE PAGES

Melin, Alexander M.; Kisner, Roger A.

2018-04-03

Improving nuclear reactor power system designs and fuel-processing technologies for safer and more efficient operation requires the development of new component designs. In particular, many of the advanced reactor designs such as the molten salt reactors and high-temperature gas-cooled reactors have operating environments beyond the capability of most currently available commercial components. To address this gap, new cross-cutting technologies need to be developed that will enable design, fabrication, and reliable operation of new classes of reactor components. The Advanced Sensor Initiative of the Nuclear Energy Enabling Technologies initiative is investigating advanced sensor and control designs that are capable of operatingmore » in these extreme environments. Under this initiative, Oak Ridge National Laboratory (ORNL) has been developing embedded instrumentation and control (I&C) for extreme environments. To develop, test, and validate these new sensing and control techniques, ORNL is building a pump test bed that utilizes submerged magnetic bearings to levitate the shaft. The eventual goal is to apply these techniques to a high-temperature (700°C) canned rotor pump that utilizes active magnetic bearings to eliminate the need for mechanical bearings and seals. The technologies will benefit the Next Generation Power Plant, Advanced Reactor Concepts, and Small Modular Reactor programs. In this paper, we will detail the design and analysis of the embedded I&C test bed with submerged magnetic bearings, focusing on the interplay between the different major systems. Then we will analyze the forces on the shaft and their role in the magnetic bearing design. Next, we will develop the radial and thrust bearing geometries needed to meet the operational requirements of the test bed. In conclusion, we will present some initial system identification results to validate the theoretical models of the test bed dynamics.« less

Design and Analysis of Embedded I&C for a Fully Submerged Magnetically Suspended Impeller Pump

DOE Office of Scientific and Technical Information (OSTI.GOV)

Melin, Alexander M.; Kisner, Roger A.

Improving nuclear reactor power system designs and fuel-processing technologies for safer and more efficient operation requires the development of new component designs. In particular, many of the advanced reactor designs such as the molten salt reactors and high-temperature gas-cooled reactors have operating environments beyond the capability of most currently available commercial components. To address this gap, new cross-cutting technologies need to be developed that will enable design, fabrication, and reliable operation of new classes of reactor components. The Advanced Sensor Initiative of the Nuclear Energy Enabling Technologies initiative is investigating advanced sensor and control designs that are capable of operatingmore » in these extreme environments. Under this initiative, Oak Ridge National Laboratory (ORNL) has been developing embedded instrumentation and control (I&C) for extreme environments. To develop, test, and validate these new sensing and control techniques, ORNL is building a pump test bed that utilizes submerged magnetic bearings to levitate the shaft. The eventual goal is to apply these techniques to a high-temperature (700°C) canned rotor pump that utilizes active magnetic bearings to eliminate the need for mechanical bearings and seals. The technologies will benefit the Next Generation Power Plant, Advanced Reactor Concepts, and Small Modular Reactor programs. In this paper, we will detail the design and analysis of the embedded I&C test bed with submerged magnetic bearings, focusing on the interplay between the different major systems. Then we will analyze the forces on the shaft and their role in the magnetic bearing design. Next, we will develop the radial and thrust bearing geometries needed to meet the operational requirements of the test bed. In conclusion, we will present some initial system identification results to validate the theoretical models of the test bed dynamics.« less

Transcript of proceedings: National Aeronautics and Space Administration, Goddard Space Flight Center, 1972 GSFC Battery Workshop, first day

NASA Technical Reports Server (NTRS)

1972-01-01

The proceedings of the 1972 NASA/Goddard Battery Workshop are reported. Topics discussed include: separators, materials and processing, test and storage experience, and improved energy density systems.

40 CFR 791.20 - Initiation of reimbursement proceeding.

Code of Federal Regulations, 2011 CFR

2011-07-01

... proceeding. (a) When persons subject to a test rule are unable to reach an agreement on the amount or method... is sought or who are seeking reimbursement, a brief description of the attempts to reach agreement...

40 CFR 791.20 - Initiation of reimbursement proceeding.

Code of Federal Regulations, 2010 CFR

2010-07-01

... proceeding. (a) When persons subject to a test rule are unable to reach an agreement on the amount or method... is sought or who are seeking reimbursement, a brief description of the attempts to reach agreement...

PBF Reactor Building (PER620). Fuel rod test assembly is on ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PBF Reactor Building (PER-620). Fuel rod test assembly is on display at PBF. Date: 1982. INEEL negative no. 82-4893 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

MTR BASEMENT. GENERAL ELECTRIC CONTROL CONSOLE FOR AIRCRAFT NUCLEAR PROPULSION ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

MTR BASEMENT. GENERAL ELECTRIC CONTROL CONSOLE FOR AIRCRAFT NUCLEAR PROPULSION EXPERIMENT NO. 1. INL NEGATIVE NO. 6510. Unknown Photographer, 9/29/1959 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

PRECAST CONCRETE WALL PANELS ARE LIFTED INTO PLACE ON MTR ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PRECAST CONCRETE WALL PANELS ARE LIFTED INTO PLACE ON MTR STEEL FRAME STRUCTURE. INL NEGATIVE NO. 1330. Unknown Photographer, 1/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

FAST CHOPPER BUILDING, TRA665, INTERIOR. UPPER LEVEL. CONCRETE WALLS. INL ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

FAST CHOPPER BUILDING, TRA-665, INTERIOR. UPPER LEVEL. CONCRETE WALLS. INL NEGATIVE NO. HD42-2. Mike Crane, Photographer, 3/2004 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

A Reload and Startup Plan for and #8233;Conversion of the NIST Research Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Diamond, D. J.; Varuttamaseni, A.

The National Institute of Standards and Technology operates a 20 MW research reactor for neutron-based research. The heavy-water moderated and cooled reactor is fueled with high-enriched uranium (HEU) but a program to convert the reactor to low-enriched uranium (LEU) fuel is underway. Among other requirements, a reload and startup test plan must be submitted to the U.S. Nuclear Regulatory Commission (NRC) for their approval. The NRC provides guidance for what should be in the plan to ensure that the licensee has sufficient information to operate the reactor safely. Hence, a plan has been generated consisting of two parts.The reload portionmore » of the plan specifies the fuel management whereby initially only two LEU fuel elements are in the core for eight fuel cycles. This is repeated until a point when the optimum approach is to place four fresh LEU elements into the reactor each cycle. This final transition is repeated and after eight cycles the reactor is completely fueled with LEU. By only adding two LEU fuel elements initially, the plan allows for the consumption of HEU fuel elements that are expected to be in storage at the time of conversion and provides additional qualification of production LEU fuel under actual operating conditions. Because the reload is to take place over many fuel cycles, startup tests will be done at different stages of the conversion. The tests, to be compared with calculations to show that the reactor will operate as planned, are the measurement of critical shim arm position and shim arm and regulating rod reactivity worths. An acceptance criterion for each test is specified based on technical specifications that relate to safe operation. Additional tests are being considered that have less safety significance but may be of interest to bolster the validation of analysis tools.« less

A reload and startup plan for conversion of the NIST research reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

D. J. Diamond

The National Institute of Standards and Technology operates a 20 MW research reactor for neutron-based research. The heavy-water moderated and cooled reactor is fueled with high-enriched uranium (HEU) but a program to convert the reactor to low-enriched uranium (LEU) fuel is underway. Among other requirements, a reload and startup test plan must be submitted to the U.S. Nuclear Regulatory Commission (NRC) for their approval. The NRC provides guidance for what should be in the plan to ensure that the licensee has sufficient information to operate the reactor safely. Hence, a plan has been generated consisting of two parts. The reloadmore » portion of the plan specifies the fuel management whereby initially only two LEU fuel elements are in the core for eight fuel cycles. This is repeated until a point when the optimum approach is to place four fresh LEU elements into the reactor each cycle. This final transition is repeated and after eight cycles the reactor is completely fueled with LEU. By only adding two LEU fuel elements initially, the plan allows for the consumption of HEU fuel elements that are expected to be in storage at the time of conversion and provides additional qualification of production LEU fuel under actual operating conditions. Because the reload is to take place over many fuel cycles, startup tests will be done at different stages of the conversion. The tests, to be compared with calculations to show that the reactor will operate as planned, are the measurement of critical shim arm position and shim arm and regulating rod reactivity worths. An acceptance criterion for each test is specified based on technical specifications that relate to safe operation. Additional tests are being considered that have less safety significance but may be of interest to bolster the validation of analysis tools.« less

10 CFR 50.58 - Hearings and report of the Advisory Committee on Reactor Safeguards.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Hearings and report of the Advisory Committee on Reactor... Hearings and report of the Advisory Committee on Reactor Safeguards. (a) Each application for a....22, or for a testing facility, shall be referred to the Advisory Committee on Reactor Safeguards for...

10 CFR 50.58 - Hearings and report of the Advisory Committee on Reactor Safeguards.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Hearings and report of the Advisory Committee on Reactor... Hearings and report of the Advisory Committee on Reactor Safeguards. (a) Each application for a....22, or for a testing facility, shall be referred to the Advisory Committee on Reactor Safeguards for...

Cost-Effective Systems for Atomic Layer Deposition

ERIC Educational Resources Information Center

Lubitz, Michael; Medina, Phillip A., IV; Antic, Aleks; Rosin, Joseph T.; Fahlman, Bradley D.

2014-01-01

Herein, we describe the design and testing of two different home-built atomic layer deposition (ALD) systems for the growth of thin films with sub-monolayer control over film thickness. The first reactor is a horizontally aligned hot-walled reactor with a vacuum purging system. The second reactor is a vertically aligned cold-walled reactor with a…

Simulation of Watts Bar Unit 1 Initial Startup Tests with Continuous Energy Monte Carlo Methods

DOE Office of Scientific and Technical Information (OSTI.GOV)

Godfrey, Andrew T; Gehin, Jess C; Bekar, Kursat B

2014-01-01

The Consortium for Advanced Simulation of Light Water Reactors* is developing a collection of methods and software products known as VERA, the Virtual Environment for Reactor Applications. One component of the testing and validation plan for VERA is comparison of neutronics results to a set of continuous energy Monte Carlo solutions for a range of pressurized water reactor geometries using the SCALE component KENO-VI developed by Oak Ridge National Laboratory. Recent improvements in data, methods, and parallelism have enabled KENO, previously utilized predominately as a criticality safety code, to demonstrate excellent capability and performance for reactor physics applications. The highlymore » detailed and rigorous KENO solutions provide a reliable nu-meric reference for VERAneutronics and also demonstrate the most accurate predictions achievable by modeling and simulations tools for comparison to operating plant data. This paper demonstrates the performance of KENO-VI for the Watts Bar Unit 1 Cycle 1 zero power physics tests, including reactor criticality, control rod worths, and isothermal temperature coefficients.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pauzi, Anas Muhamad, E-mail: Anas@uniten.edu.my; Cioncolini, Andrea; Iacovides, Hector

The Molten Salt Reactor (MSR) is one of the most promising system proposed by Generation IV Forum (GIF) for future nuclear reactor systems. Advantages of the MSR are significantly larger compared to other reactor system, and is mainly achieved from its liquid nature of fuel and coolant. Further improvement to this system, which is a natural circulating molten fuel salt inside its tube in the reactor core is proposed, to achieve advantages of reducing and simplifying the MSR design proposed by GIF. Thermal hydraulic analysis on the proposed system was completed using a commercial computation fluid dynamics (CFD) software calledmore » FLUENT by ANSYS Inc. An understanding on theory behind this unique natural circulation flow inside the tube caused by fission heat generated in molten fuel salt and tube cooling was briefly introduced. Currently, no commercial CFD software could perfectly simulate natural circulation flow, hence, modeling this flow problem in FLUENT is introduced and analyzed to obtain best simulation results. Results obtained demonstrate the existence of periodical transient nature of flow problem, hence improvements in tube design is proposed based on the analysis on temperature and velocity profile. Results show that the proposed system could operate at up to 750MW core power, given that turbulence are enhanced throughout flow region, and precise molten fuel salt physical properties could be defined. At the request of the authors and the Proceedings Editor the name of the co-author Andrea Cioncolini was corrected from Andrea Coincolini. The same name correction was made in the Acknowledgement section on page 030004-10 and in reference number 4. The updated article was published on 11 May 2015.« less

Risk factors for bovine tuberculosis in low incidence regions related to the movements of cattle.

PubMed

Gates, M Carolyn; Volkova, Victoriya V; Woolhouse, Mark E J

2013-11-09

Bovine tuberculosis (bTB) remains difficult to eradicate from low incidence regions partly due to the imperfect sensitivity and specificity of routine intradermal tuberculin testing. Herds with unconfirmed reactors that are incorrectly classified as bTB-negative may be at risk of spreading disease, while those that are incorrectly classified as bTB-positive may be subject to costly disease eradication measures. This analysis used data from Scotland in the period leading to Officially Tuberculosis Free recognition (1) to investigate the risks associated with the movements of cattle from herds with different bTB risk classifications and (2) to identify herd demographic characteristics that may aid in the interpretation of tuberculin testing results. From 2002 to 2009, for every herd with confirmed bTB positive cattle identified through routine herd testing, there was an average of 2.8 herds with at least one unconfirmed positive reactor and 18.9 herds with unconfirmed inconclusive reactors. Approximately 75% of confirmed bTB positive herds were detected through cattle with no known movements outside Scotland. At the animal level, cattle that were purchased from Scottish herds with unconfirmed positive reactors and a recent history importing cattle from endemic bTB regions were significantly more likely to react positively on routine intradermal tuberculin tests, while cattle purchased from Scottish herds with unconfirmed inconclusive reactors were significantly more likely to react inconclusively. Case-case comparisons revealed few demographic differences between herds with confirmed positive, unconfirmed positive, and unconfirmed inconclusive reactors, which highlights the difficulty in determining the true disease status of herds with unconfirmed tuberculin reactors. Overall, the risk of identifying reactors through routine surveillance decreased significantly over time, which may be partly attributable to changes in movement testing regulations and the volume of cattle imported from endemic regions. Although the most likely source of bTB infections in Scotland was cattle previously imported from endemic regions, we found indirect evidence of transmission within Scottish cattle farms and cannot rule out the possibility of low level transmission between farms. Further investigation is needed to determine whether targeting herds with unconfirmed reactors and a history of importing cattle from high risk regions would benefit control efforts.

THE EXPERIENCE IN THE UNITED STATES WITH REACTOR OPERATION AND REACTOR SAFEGUARDS

DOE Office of Scientific and Technical Information (OSTI.GOV)

McCullough, C.R.

1958-10-31

Reactors are operating or planned at locations in the United States in cities, near cities, and at remote locations. There is a general pattern that the higher power reactors are not in, but fairly uear cities, and the testing reactors for more hazardous experiments are at remote locations. A great deal has been done on the theoretical and experimental study of importunt features of reactor design. The metal-water reaction is still a theoretical possibility but tests of fuel element burnout under conditions approaching reactor operation gave no reaction. It appears that nucleate boiling does not necessarily result in steam blanketingmore » and fuel melting. Much attention is being given to the calculation of core kinetics but it is being found that temperature, power, and void coefficients cannot be calculated with accuracy and experiments are required. Some surprises are found giving positive localized void coefficients. Possible oscillatory behavior of reactors is being given careful study. No dangerous oscillations have been found in operating reactors but osciliations hare appeared in experimeats. The design of control and safety systems varies wvith different constructors. The relation of control to the kinetic behavior of the reactor is being studied. The importance of sensing element locations in order to know actual local reactor power level is being recognized. The time constants of instrumentation as related to reactor kinetics are being studied. Pressure vessels for reactors are being designed and manufactured. Many of these are beyond any previous experience. The stress problem is being given careful study. The effect of radiation is being studied experimentally. The stress problems of piping and pressure vessels is a difficult design problem being met successfully in reactor plants. The proper organization and procedure for operation of reactors is being evolved for resourch, testing, and power reactors. The importance of written standards and instructions for both normal and abnormal operating conditions is recogmized. Corfinement of radioactive materials either by tight steel shells, tight buildings, or semi-tight structures vented through filters is considered necessary in the United States. A discussion will be given of specifications, construction, and testing of these structures. The need for emergency plans has been stressed by recent experiences in radioactive releases. The problems of such plans to cover all grades of accidents will be discussed. The theoretical consequences of releases of radioactive materials have been studied and these results will be compared with actual experience. The problem of exposures from normal and abnormal operetion of reactors is a problem of desiga and operation on one hand and the amount of damage to be expected on the other. The safeguard problem is closely related to the acceptable doses of radiouctivity which the ICRP recommend. The future of atomic energy depends upon adequate safeguards and economical design and operation. Accepted criteria are required to guide designers as to the proper balance of caution and boldness. (auth)« less

Guideline for Performing Systematic Approach to Evaluate and Qualify Legacy Documents that Support Advanced Reactor Technology Activity

DOE Office of Scientific and Technical Information (OSTI.GOV)

Honma, George

The establishment of a systematic process for the evaluation of historic technology information for use in advanced reactor licensing is described. Efforts are underway to recover and preserve Experimental Breeder Reactor II and Fast Flux Test Facility historical data. These efforts have generally emphasized preserving information from data-acquisition systems and hard-copy reports and entering it into modern electronic formats suitable for data retrieval and examination. The guidance contained in this document has been developed to facilitate consistent and systematic evaluation processes relating to quality attributes of historic technical information (with focus on sodium-cooled fast reactor (SFR) technology) that will bemore » used to eventually support licensing of advanced reactor designs. The historical information may include, but is not limited to, design documents for SFRs, research-and-development (R&D) data and associated documents, test plans and associated protocols, operations and test data, international research data, technical reports, and information associated with past U.S. Nuclear Regulatory Commission (NRC) reviews of SFR designs. The evaluation process is prescribed in terms of SFR technology, but the process can be used to evaluate historical information for any type of advanced reactor technology. An appendix provides a discussion of typical issues that should be considered when evaluating and qualifying historical information for advanced reactor technology fuel and source terms, based on current light water reactor (LWR) requirements and recent experience gained from Next Generation Nuclear Plant (NGNP).« less

77 FR 64148 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Regulatory...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-18

... Regulatory Guides (RG) RG 1.79, ````Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors,'' Revision 2 and RG 1.79.1, ``Initial Test Program of Emergency Core Cooling Systems for...

A 640 foot per second impact test of a two foot diameter model nuclear reactor containment system without fracture

NASA Technical Reports Server (NTRS)

Puthoff, R. L.

1971-01-01

An impact test was conducted on an 1142 pound 2 foot diameter sphere model. The purpose of this test was to determine the feasibility of containing the fission products of a mobile reactor in an impact. The model simulated the reactor core, energy absorbing gamma shielding, neutron shielding and the containment vessel. It was impacted against an 18,000 pound reinforced concrete block. The model was significantly deformed and the concrete block demolished. No leaks were detected nor cracks observed in the model after impact.

High velocity continuous-flow reactor for the production of solar grade silicon

NASA Technical Reports Server (NTRS)

Woerner, L.

1977-01-01

The feasibility of a high volume, high velocity continuous reduction reactor as an economical means of producing solar grade silicon was tested. Bromosilanes and hydrogen were used as the feedstocks for the reactor along with preheated silicon particles which function both as nucleation and deposition sites. A complete reactor system was designed and fabricated. Initial preheating studies have shown the stability of tetrabromosilane to being heated as well as the ability to preheat hydrogen to the desired temperature range. Several test runs were made and some silicon was obtained from runs carried out at temperatures in excess of 1180 K.

Wide-range structurally optimized channel for monitoring the certified power of small-core reactors

NASA Astrophysics Data System (ADS)

Koshelev, A. S.; Kovshov, K. N.; Ovchinnikov, M. A.; Pikulina, G. N.; Sokolov, A. B.

2016-12-01

The results of tests of a prototype version of a channel for monitoring the certified power of small-core reactors performed at the BR-K1 reactor at the All-Russian Scientific Research Institute of Experimental Physics are reported. An SNM-11 counter and commercial KNK-4 and KNK-3 compensated ion chambers were used as neutron detectors in the tested channel, and certified NCMM and CCMM measurement modules controlled by a PC with specialized software were used as measuring instruments. The specifics of metrological assurance of calibration of the channel in the framework of reactor power monitoring are discussed.

[Radiation ecological environment in the Republic of Kazakhstan in the vicinity of the reactors and on the territory of the Semipalatinsk Test Site].

PubMed

Kim, D S

2012-01-01

The results of research into the environmental conditions in the regions of location of the pressurized water reactor WWR-K, fast neutron breeder BN-350 and on the territory of the Semipalatinsk Test Site are represented. The effects of the exposure to aerosol emissions from WWR-K and BN-350 reactors on the environment are summarized. We present some arguments in favor of the safe operation of fission reactors in compliance with the rules and norms of nuclear and radiation protection and the efficient disposal of radioactive waste on the territory of the Republic.

Wide-range structurally optimized channel for monitoring the certified power of small-core reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Koshelev, A. S., E-mail: alexsander.coshelev@yandex.ru; Kovshov, K. N.; Ovchinnikov, M. A.

The results of tests of a prototype version of a channel for monitoring the certified power of small-core reactors performed at the BR-K1 reactor at the All-Russian Scientific Research Institute of Experimental Physics are reported. An SNM-11 counter and commercial KNK-4 and KNK-3 compensated ion chambers were used as neutron detectors in the tested channel, and certified NCMM and CCMM measurement modules controlled by a PC with specialized software were used as measuring instruments. The specifics of metrological assurance of calibration of the channel in the framework of reactor power monitoring are discussed.

Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jennifer Lyons; Wade R. Marcum; Mark D. DeHart

2014-01-01

The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by themore » Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.« less

Corrosion resistance of 0Kh18N10T steel in gadolinium nitrate solutions in the liquid regulation of the reactivity of nuclear reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ganzha, V.D.; Konoplev, K.A.; Mashchetov, V.P.

1986-03-01

This study was carried out in connection with the preparation of the design for the PIK research reactor. The corrosion resistance of 0Kh18N10T steel in gadolinium nitrate solutions was tested in laboratory, ampule, and loop corrosion tests. At all stages of the tests, the authors investigated the effect produced on the corrosion processes by factors related to the technology of preparation of the equipment (mechanical working of the surfaces, welding, sensitizing, annealing, stressed state of the material, cracks, etc.). Ampule tests were conducted in order to determine the effect produced by reactor radiation and shutdown regimes on the corrosion resistancemore » of the steel. Special ampules made of 0Kh18N10T steel were filled with gadolinium nitrate solutions of various concentrations, sealed, and irradiated for a long period in the core of the VVR-M reactor at a temperature of 20-50 degrees C. The results of the tests are shown. The investigations showed that the corrosion of 0Kh18N10T steel in solutions of gadolinium nitrate is uniform, regardless of the state of the surface, the concentration of gadolinium nitrate, the duration of the tests, the action of the reactor radiation under static and dynamic conditions, and the presence of mechanical stresses.« less

United States and Russian Cooperation on Issues of Nuclear Nonproliferation

DTIC Science & Technology

2005-06-01

Reactors ( RERTR ) This project works with Russia to facilitate conversion of its research and test reactors from highly enriched uranium (HEU) fuel...reactor fuel purchase, accelerated RERTR activities, and accelerated Material Conversion and Consolidation implementation. 89 j. Fissile Materials

IET. Typical detail during Snaptran reactor experiments. Shielding bricks protect ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

IET. Typical detail during Snaptran reactor experiments. Shielding bricks protect ion chamber beneath reactor on dolly. Photographer: Page Comiskey. Date: August 11, 1965. INEEL negative no. 65-4039 - Idaho National Engineering Laboratory, Test Area North, Scoville, Butte County, ID

Design, Construction and Testing of an In-Pile Loop for PWR (Pressurized Water Reactor) Simulation.

DTIC Science & Technology

1987-06-01

computer modeling remains at best semiempirical (C-i), this large variation in scaling factor makes extrapolation of data impossible. The DIDO Water...in a full scale PWR are not practical. The reactor plant is not controlled to tolerances necessary for research, and utilities are reluctant to vary...MIT Reactor Safeguards Committee, in revision 1 to the PCCL Safety Evaluation Report (SER), for final approval to begin in-pile testing and

The International Experimental Thermal Hydraulic Systems database – TIETHYS: A new NEA validation tool

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rohatgi, Upendra S.

Nuclear reactor codes require validation with appropriate data representing the plant for specific scenarios. The thermal-hydraulic data is scattered in different locations and in different formats. Some of the data is in danger of being lost. A relational database is being developed to organize the international thermal hydraulic test data for various reactor concepts and different scenarios. At the reactor system level, that data is organized to include separate effect tests and integral effect tests for specific scenarios and corresponding phenomena. The database relies on the phenomena identification sections of expert developed PIRTs. The database will provide a summary ofmore » appropriate data, review of facility information, test description, instrumentation, references for the experimental data and some examples of application of the data for validation. The current database platform includes scenarios for PWR, BWR, VVER, and specific benchmarks for CFD modelling data and is to be expanded to include references for molten salt reactors. There are place holders for high temperature gas cooled reactors, CANDU and liquid metal reactors. This relational database is called The International Experimental Thermal Hydraulic Systems (TIETHYS) database and currently resides at Nuclear Energy Agency (NEA) of the OECD and is freely open to public access. Going forward the database will be extended to include additional links and data as they become available. https://www.oecd-nea.org/tiethysweb/« less

STUDIES OF FAST REACTOR FUEL ELEMENT BEHAVIOR UNDER TRANSIENT HEATING TO FAILURE. I. INITIAL EXPERIMENTS ON METALLIC SAMPLES IN THE ABSENCE OF COOLANT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dickerman, C. E.; Sowa, E. S.; Okrent, D.

1961-08-01

Meltdown tests on single metallic unirradiated fuel elements in TREAT are described. The fuel elements (EBRII Mark I fuel pins, EBR-II fuel pins with retractory Nb or Ta cladding, and Fermi-I fuel pins) are tested in an inert atmosphere, with no coolant. The fuel elements are exposed to reactor power bursts of 200 msec to 25 sec duration, under conditions simulating fast reactor operations. For these tests, the type of power burst, the integrated power, the fuel enrichment, the maximum cladding temperature, and the effects of the test on the fuel element are recorded. ( T.F.H.)

The prototype fast reactor at Dounreay, Scotland. Process and engineering development for sodium removal

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mann, A.; Herrick, R.; Gunn, J.

2007-07-01

Dounreay was home to commercial fast reactor development in the UK. Following the construction and operation of the Dounreay Fast Reactor, a sodium-cooled Prototype Fast Reactor (PFR), was constructed. PFR started operating in 1974, closed in 1994 and is presently being decommissioned. To date the bulk of the sodium has been removed and treated. Due to the design of the existing extraction system however, a sodium pool will remain in the heel of the reactor. To remove this sodium, a pump/camera system was developed, tested and deployed. The Water Vapour Nitrogen (WVN) process has been selected to allow removal ofmore » the final sodium residues from the reactor. Due to the design of the reactor and potential for structural damage should Normal WVN (which produces hydrated sodium hydroxide) be used, Low Concentration WVN (LC WVN) has been developed. Pilot scale testing has shown that it is possible treat the reactor within 18 months at a WVN concentration of up to 4% v/v and temperature of 120 deg. C. At present the equipment that will be used to apply LC WVN to the reactor is being developed at the detail design stage. and is expected to be deployed within the next few years. (authors)« less

Tower Shielding Reactor II design and operation report: Vol. 2. Safety Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Holland, L. B.; Kolb, J. O.

1970-01-01

Information on the Tower Shielding Reactor II is contained in the TSR-II Design and Operation Report and in the Tower Shielding Facility Manual. The TSR-II Design and Operating Report consists of three volumes. Volume 1 is Descriptions of the Tower Shielding Reactor II and Facility; Volume 2 is Safety analysis of the Tower Shielding Reactor II; and Volume 3 is the Assembly and Testing of the Tower Shielding Reactor II Control Mechanism Housing.

ADVANCED HEAT TRANSFER TEST FACILITY, TRA666A. ELEVATIONS. ROOF FRAMING PLAN. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ADVANCED HEAT TRANSFER TEST FACILITY, TRA-666A. ELEVATIONS. ROOF FRAMING PLAN. CONCRETE BLOCK SIDING. SLOPED ROOF. ROLL-UP DOOR. AIR INTAKE ENCLOSURE ON NORTH SIDE. F.C. TORKELSON 842-MTR-666-A5, 8/1966. INL INDEX NO. 531-0666-00-851-152258, REV. 2. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Fuel processing in integrated micro-structured heat-exchanger reactors

NASA Astrophysics Data System (ADS)

Kolb, G.; Schürer, J.; Tiemann, D.; Wichert, M.; Zapf, R.; Hessel, V.; Löwe, H.

Micro-structured fuel processors are under development at IMM for different fuels such as methanol, ethanol, propane/butane (LPG), gasoline and diesel. The target application are mobile, portable and small scale stationary auxiliary power units (APU) based upon fuel cell technology. The key feature of the systems is an integrated plate heat-exchanger technology which allows for the thermal integration of several functions in a single device. Steam reforming may be coupled with catalytic combustion in separate flow paths of a heat-exchanger. Reactors and complete fuel processors are tested up to the size range of 5 kW power output of a corresponding fuel cell. On top of reactor and system prototyping and testing, catalyst coatings are under development at IMM for numerous reactions such as steam reforming of LPG, ethanol and methanol, catalytic combustion of LPG and methanol, and for CO clean-up reactions, namely water-gas shift, methanation and the preferential oxidation of carbon monoxide. These catalysts are investigated in specially developed testing reactors. In selected cases 1000 h stability testing is performed on catalyst coatings at weight hourly space velocities, which are sufficiently high to meet the demands of future fuel processing reactors.

Further Development of Crack Growth Detection Techniques for US Test and Research Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kohse, Gordon; Carpenter, David M.; Ostrovsky, Yakov

One of the key issues facing Light Water Reactors (LWRs) in extending lifetimes beyond 60 years is characterizing the combined effect of irradiation and water chemistry on material degradation and failure. Irradiation Assisted Stress Corrosion Cracking (IASCC), in which a crack propagates in a susceptible material under stress in an aggressive environment, is a mechanism of particular concern. Full understanding of IASCC depends on real time crack growth data acquired under relevant irradiation conditions. Techniques to measure crack growth in actively loaded samples under irradiation have been developed outside the US - at the Halden Boiling Water Reactor, for example.more » Several types of IASCC tests have also been deployed at the MITR, including passively loaded crack growth measurements and actively loaded slow strain rate tests. However, there is not currently a facility available in the US to measure crack growth on actively loaded, pre-cracked specimens in LWR irradiation environments. A joint program between the Idaho National Laboratory (INL) and the Massachusetts Institute of Technology (MIT) Nuclear Reactor Laboratory (NRL) is currently underway to develop and demonstrate such a capability for US test and research reactors. Based on the Halden design, the samples will be loaded using miniature high pressure bellows and a compact loading mechanism, with crack length measured in real time using the switched Direct Current Potential Drop (DCPD) method. The basic design and initial mechanical testing of the load system and implementation of the DCPD method have been previously reported. This paper presents the results of initial autoclave testing at INL and the adaptation of the design for use in the high pressure, high temperature water loop at the MITR 6 MW research reactor, where an initial demonstration is planned in mid-2015. Materials considerations for the high pressure bellows are addressed. Design modifications to the loading mechanism required by the size constraints of the MITR water loop are described. The safety case for operation of the high pressure gas-driven bellows mechanism is also presented. Key issues are the design and response of systems to limit gas flow in the event of a high pressure gas leak in the in-core autoclave. Integrity of the autoclave must be maintained and reactivity effects due to voiding of the loop coolant must be shown to be within the reactor technical specifications. The technical development of the crack growth monitor for application in the INL Advanced Test Reactor or the MITR can act as a template for adaptation of this technology in other reactors. (authors)« less

MTR CONTROL ROOM WITH CONTROL CONSOLE AND STATUS READOUTS ALONG ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

MTR CONTROL ROOM WITH CONTROL CONSOLE AND STATUS READOUTS ALONG WALL. WORKERS MAKE ELECTRICAL AND OTHER CONNECTIONS. INL NEGATIVE NO. 4289. Unknown Photographer, 2/26/1952 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

10 CFR 100.1 - Purpose.

Code of Federal Regulations, 2010 CFR

2010-01-01

... approval requirements for proposed sites for stationary power and testing reactors subject to part 50 or part 52 of this chapter. (b) There exists a substantial base of knowledge regarding power reactor... approach incorporates the appropriate standards and criteria for approval of stationary power and testing...

SOUTH WING, TRA661. SOUTH SIDE. CAMERA FACING NORTH. MTR HIGH ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

SOUTH WING, TRA-661. SOUTH SIDE. CAMERA FACING NORTH. MTR HIGH BAY BEYOND. INL NEGATIVE NO. HD46-45-3. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR HEAT EXCHANGER BUILDING, TRA644. WORKERS ARE INSTALLING HEAT EXCHANGER ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR HEAT EXCHANGER BUILDING, TRA-644. WORKERS ARE INSTALLING HEAT EXCHANGER PIPING. INL NEGATIVE NO. 56-3122. Jack L. Anderson, Photographer, 9/21/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

MTR AND ETR COMPLEXES. CAMERA FACING EASTERLY TOWARD CHEMICAL PROCESSING ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

MTR AND ETR COMPLEXES. CAMERA FACING EASTERLY TOWARD CHEMICAL PROCESSING PLANT. MTR AND ITS ATTACHMENTS IN FOREGROUND. ETR BEYOND TO RIGHT. INL NEGATIVE NO. 56-4100. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR COMPRESSOR BUILDING, TRA643. COMPRESSORS AND OTHER EQUIPMENT INSTALLED. METAL ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR COMPRESSOR BUILDING, TRA-643. COMPRESSORS AND OTHER EQUIPMENT INSTALLED. METAL ROOF AND CONCRETE BLOCK WALLS. INL NEGATIVE NO. 61-4536. Unknown Photographer, ca. 1961. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

State-Mandated Student Assessment Programs: An Examination of the Issues. Proceedings of a Teleconference (October 3, 1985) and a Conference of National and State Educators (Stillwater, Oklahoma, October 4, 1985).

ERIC Educational Resources Information Center

Perry, Katye M., Ed.; And Others

These combined proceedings summarize presentations of two conference programs designed to address state-mandated student achievement issues. Dr. Terrel Bell delivered the keynote speech at a teleconference focusing on a broad perspective of competency testing. He discussed testing problems, as well as positive aspects such as comparison, diagnosis…

Development of variable-width ribbon heating elements for liquid-metal and gas-cooled fast breeder reactor fuel-pin simulators

DOE Office of Scientific and Technical Information (OSTI.GOV)

McCulloch, R.W.; Post, D.W.; Lovell, R.T.

1981-04-01

Variable-width ribbon heating elements that provide a chopped-cosine variable heat flux profile have been fabricated for fuel pin simulators used in test loops by the Breeder Reactor Program Thermal-Hydraulic Out-of-Reactor Safety test facility and the Gas-Cooled Fast Breeder Reactor-Core Flow Test Loop. Thermal, mechanical, and electrical design considerations are used to derive an analytical expression that precisely describes ribbon contour in terms of the major fabrication parameters. These parameters are used to generate numerical control tapes that control ribbon cutting and winding machines. Infrared scanning techniques are developed to determine the optimum transient thermal profile of the coils and relatemore » this profile to that generated by the coils in completed fuel pin simulators.« less

10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

Code of Federal Regulations, 2011 CFR

2011-01-01

.... Containment inspection. B. Repordkeeping of test results. I. Introduction One of the conditions of all... following: A. Type A test—1. Pretest requirements. (a) Containment inspection in accordance with V. A. shall.... During the period between the completion of one Type A test and the initiation of the containment...

PBF Reactor Building (PER620). Detail of fuel test assembly in ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PBF Reactor Building (PER-620). Detail of fuel test assembly in preparation for test. When complete, it will fit into in-pile tube. The maximum outside diameter of which must be about 8.25 inches. Date: 1982. INEEL negative no. 82-4908 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

78 FR 33132 - Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-03

... Fuel Elements for Use in Research and Test Reactors AGENCY: Nuclear Regulatory Commission. ACTION... Research and Test Reactors.'' This guide describes a method that the staff of the NRC considers acceptable... assurance program for verifying the quality of plate-type uranium-aluminum fuel elements used in research...

Imaging Fukushima Daiichi reactors with muons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Miyadera, Haruo; Borozdin, Konstantin N.; Greene, Steve J.

2013-05-15

A study of imaging the Fukushima Daiichi reactors with cosmic-ray muons to assess the damage to the reactors is presented. Muon scattering imaging has high sensitivity for detecting uranium fuel and debris even through thick concrete walls and a reactor pressure vessel. Technical demonstrations using a reactor mockup, detector radiation test at Fukushima Daiichi, and simulation studies have been carried out. These studies establish feasibility for the reactor imaging. A few months of measurement will reveal the spatial distribution of the reactor fuel. The muon scattering technique would be the best and probably the only way for Fukushima Daiichi tomore » make this determination in the near future.« less

Imaging Fukushima Daiichi reactors with muons

NASA Astrophysics Data System (ADS)

Miyadera, Haruo; Borozdin, Konstantin N.; Greene, Steve J.; Lukić, Zarija; Masuda, Koji; Milner, Edward C.; Morris, Christopher L.; Perry, John O.

2013-05-01

A study of imaging the Fukushima Daiichi reactors with cosmic-ray muons to assess the damage to the reactors is presented. Muon scattering imaging has high sensitivity for detecting uranium fuel and debris even through thick concrete walls and a reactor pressure vessel. Technical demonstrations using a reactor mockup, detector radiation test at Fukushima Daiichi, and simulation studies have been carried out. These studies establish feasibility for the reactor imaging. A few months of measurement will reveal the spatial distribution of the reactor fuel. The muon scattering technique would be the best and probably the only way for Fukushima Daiichi to make this determination in the near future.

Space Nuclear Reactor Engineering

DOE Office of Scientific and Technical Information (OSTI.GOV)

Poston, David Irvin

We needed to find a space reactor concept that could be attractive to NASA for flight and proven with a rapid turnaround, low-cost nuclear test. Heat-pipe-cooled reactors coupled to Stirling engines long identified as the easiest path to near-term, low-cost concept.

Application of ATHLET/DYN3D coupled codes system for fast liquid metal cooled reactor steady state simulation

NASA Astrophysics Data System (ADS)

Ivanov, V.; Samokhin, A.; Danicheva, I.; Khrennikov, N.; Bouscuet, J.; Velkov, K.; Pasichnyk, I.

2017-01-01

In this paper the approaches used for developing of the BN-800 reactor test model and for validation of coupled neutron-physic and thermohydraulic calculations are described. Coupled codes ATHLET 3.0 (code for thermohydraulic calculations of reactor transients) and DYN3D (3-dimensional code of neutron kinetics) are used for calculations. The main calculation results of reactor steady state condition are provided. 3-D model used for neutron calculations was developed for start reactor BN-800 load. The homogeneous approach is used for description of reactor assemblies. Along with main simplifications, the main reactor BN-800 core zones are described (LEZ, MEZ, HEZ, MOX, blankets). The 3D neutron physics calculations were provided with 28-group library, which is based on estimated nuclear data ENDF/B-7.0. Neutron SCALE code was used for preparation of group constants. Nodalization hydraulic model has boundary conditions by coolant mass-flow rate for core inlet part, by pressure and enthalpy for core outlet part, which can be chosen depending on reactor state. Core inlet and outlet temperatures were chosen according to reactor nominal state. The coolant mass flow rate profiling through the core is based on reactor power distribution. The test thermohydraulic calculations made with using of developed model showed acceptable results in coolant mass flow rate distribution through the reactor core and in axial temperature and pressure distribution. The developed model will be upgraded in future for different transient analysis in metal-cooled fast reactors of BN type including reactivity transients (control rods withdrawal, stop of the main circulation pump, etc.).

Assessment of Sensor Technologies for Advanced Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Korsah, Kofi; Kisner, R. A.; Britton Jr., C. L.

This paper provides an assessment of sensor technologies and a determination of measurement needs for advanced reactors (AdvRx). It is a summary of a study performed to provide the technical basis for identifying and prioritizing research targets within the instrumentation and control (I&C) Technology Area under the Department of Energy’s (DOE’s) Advanced Reactor Technology (ART) program. The study covered two broad reactor technology categories: High Temperature Reactors and Fast Reactors. The scope of “High temperature reactors” included Gen IV reactors whose coolant exit temperatures exceed ≈650 °C and are moderated (as opposed to fast reactors). To bound the scope formore » fast reactors, this report reviewed relevant operating experience from US-operated Sodium Fast Reactor (SFR) and relevant test experience from the Fast Flux Test Facility (FFTF). For high temperature reactors the study showed that in many cases instrumentation have performed reasonably well in research and demonstration reactors. However, even in cases where the technology is “mature” (such as thermocouples), HTGRs can benefit from improved technologies. Current HTGR instrumentation is generally based on decades-old technology and adapting newer technologies could provide significant advantages. For sodium fast reactors, the study found that several key research needs arise around (1) radiation-tolerant sensor design for in-vessel or in-core applications, where possible non-invasive sensing approaches for key parameters that minimize the need to deploy sensors in-vessel, (2) approaches to exfiltrating data from in-vessel sensors while minimizing penetrations, (3) calibration of sensors in-situ, and (4) optimizing sensor placements to maximize the information content while minimizing the number of sensors needed.« less

Current and prospective safety issues at the HFBR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tichler, P.R.

The Brookhaven high-flux beam reactor (HFBR) was designed primarily to produce external neutron beams for experimental research. It is cooled, moderated, and reflected by heavy water and uses materials test reactor and engineering test reactor type of fuel elements containing enriched uranium. The reactor power when operation began in 1965 was 40 MW, was raised to 60 MW in 1982 after a number of plant modifications, and operated at that level until 1989. Since that time, safety questions have been raised that resulted in extended shutdowns and a reduction in operating power to 30 MW. This paper discusses the principalmore » safety issues and plans for their resolution and return to 60-MW operation. In addition, radiation embrittlement of the reactor vessel and thermal shield and its effect on the life of the facility are briefly discussed.« less

WATER PROCESS SYSTEM FLOW DIAGRAM FOR MTR, TRA603. SUMMARY OF ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

WATER PROCESS SYSTEM FLOW DIAGRAM FOR MTR, TRA-603. SUMMARY OF COOLANT FLOW FROM WORKING RESERVOIR TO INTERIOR OF REACTOR'S THERMAL SHIELD. NAMES TANK SECTIONS. PIPE AND DRAIN-LINE SIZES. SHOWS DIRECTION OF AIR FLOW THROUGH PEBBLE AND GRAPHITE BLOCK ZONE. NEUTRON CURTAIN AND THERMAL COLUMN DOOR. BLAW-KNOX 3150-92-7, 3/1950. INL INDEX NO. 531-0603-51-098-100036, REV. 6. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETRMTR MECHANICAL SERVICES BUILDING, TRA653. CAMERA FACING NORTHWEST AS BUILDING ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR-MTR MECHANICAL SERVICES BUILDING, TRA-653. CAMERA FACING NORTHWEST AS BUILDING WAS NEARLY COMPLETE. INL NEGATIVE NO. 57-3653. K. Mansfield, Photographer, 7/22/1957 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

DESIGN CRITERIA FOR HIGH TEMPERATURE LATTICE TEST REACTOR PROJECT CAH-100

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ballard, D.L.; Brown, W.W.; Harrison, C.W.

Design and construction specifications to be followed in the development of the reactor, its associated systems and experimental facilities, and the housing and required services for the facility are presented. The testing procedures to be used are outlined. (D.C.W.)

FAST CHOPPER BUILDING, TRA665. DETAIL OF STEEL DOOR ENTRY TO ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

FAST CHOPPER BUILDING, TRA-665. DETAIL OF STEEL DOOR ENTRY TO LOWER LEVEL. CAMERA FACING NORTH. INL NEGATIVE NO. HD42-1. Mike Crane, Photographer, 3/2004 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

PBF Reactor Building (PER620). Camera on main floor faces south ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PBF Reactor Building (PER-620). Camera on main floor faces south (open) doorway. In foreground is canal gate, lined with stainless steel and painted with protective coatings. Reactor pit is round with protective coatings. Reactor put is round form discernible beyond. Lifting beams and rigging are in place for a load test before reactor vessel arrives. Photographer: John Capek. Date: January 26, 1970. INEEL negative no. 70-347 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Pilot plant operation of a nonadiabatic methanation reactor. [15 refs. ; Raney nickel catalyst

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schehl, R.R.; Pennline, H.W.; Strakey, J.P.

The design and operation of a pilot plant scale hybrid methanation reactor is discussed. The hybrid methanator, utilizing a finned, Raney nickel coated insert, consolidates features of the tube-wall and hot-gas-recycle methanation reactors. Data are presented from four tests lasting from 3/sup 1///sub 2/ weeks to three months. Topics discussed include conversion, product yields, catalyst properties, and reactor temperature profiles. A one-dimensional mathematical model capable of explaining reactor performance trends is employed.

Hodoscope Cineradiography Of Nuclear Fuel Destruction Experiments

NASA Astrophysics Data System (ADS)

De Volpi, A.

1983-08-01

Nuclear reactor safety studies have applied cineradiographic techniques to achieve key information regarding the durability of fuel elements that are subjected to destructive transients in test reactors. Beginning with its development in 1963, the fast-neutron hodoscope has recorded data at the TREAT reactor in the United States of America. Consisting of a collimator instrumented with several hundred parallel channels of detectors and associated instrumentation, the hodoscope measures fuel motion that takes place within thick-walled steel test containers. Fuel movement is determined by detecting the emission of fast neutrons induced in the test capsule by bursts of the test reactor that last from 0.3 to 30 s. The system has been designed so as to achieve under certain typical conditions( horizontal) spatial resolution less than lmm, time resolution close to lms, mass resolution below 0.1 g, with adequate dynamic range and recording duration. A variety of imaging forms have been developed to display the results of processing and analyzing recorded data.*

Shear compression testing of glass-fibre steel specimens after 4K reactor irradiation: Present status and facility upgrade

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gerstenberg, H.; Kraehling, E.; Katheder, H.

1997-06-01

The shear strengths of various fibre reinforced resins being promising candidate insulators for superconducting coils to be used tinder a strong radiation load, e.g. in future fusion reactors were investigated prior and subsequent to reactor in-core irradiation at liquid helium temperature. A large number of sandwich-like (steel-bonded insulation-steel) specimens representing a widespread variety of materials and preparation techniques was exposed to irradiation doses of up to 5 x 10{sup 7} Gy in form of fast neutrons and {gamma}-radiation. In a systematic study several experimental parameters including irradiation dose, postirradiation storage temperature and measuring temperature were varied before the determination ofmore » the ultimate shear strength. The results obtained from the different tested materials are compared. In addition an upgrade of the in-situ test rig installed at the Munich research reactor is presented, which allows combined shear/compression loading of low temperature irradiated specimens and provides a doubling of the testing rate.« less

Application of a Virtual Reactivity Feedback Control Loop in Non-Nuclear Testing of a Fast Spectrum Reactor

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Forsbacka, Matthew

2004-01-01

For a compact, fast-spectrum reactor, reactivity feedback is dominated by core deformation at elevated temperature. Given the use of accurate deformation measurement techniques, it is possible to simulate nuclear feedback in non-nuclear electrically heated reactor tests. Implementation of simulated reactivity feedback in response to measured deflection is being tested at the NASA Marshall Space Flight Center Early Flight Fission Test Facility (EFF-TF). During tests of the SAFE-100 reactor prototype, core deflection was monitored using a high resolution camera. "virtual" reactivity feedback was accomplished by applying the results of Monte Carlo calculations (MCNPX) to core deflection measurements; the computational analysis was used to establish the reactivity worth of van'ous core deformations. The power delivered to the SAFE-100 prototype was then dusted accordingly via kinetics calculations, The work presented in this paper will demonstrate virtual reactivity feedback as core power was increased from 1 kilowatt(sub t), to 10 kilowatts(sub t), held approximately constant at 10 kilowatts (sub t), and then allowed to decrease based on the negative thermal reactivity coefficient.

Hyperthermal Environments Simulator for Nuclear Rocket Engine Development

NASA Technical Reports Server (NTRS)

Litchford, Ron J.; Foote, John P.; Clifton, W. B.; Hickman, Robert R.; Wang, Ten-See; Dobson, Christopher C.

2011-01-01

An arc-heater driven hyperthermal convective environments simulator was recently developed and commissioned for long duration hot hydrogen exposure of nuclear thermal rocket materials. This newly established non-nuclear testing capability uses a high-power, multi-gas, wall-stabilized constricted arc-heater to produce hightemperature pressurized hydrogen flows representative of nuclear reactor core environments, excepting radiation effects, and is intended to serve as a low-cost facility for supporting non-nuclear developmental testing of hightemperature fissile fuels and structural materials. The resulting reactor environments simulator represents a valuable addition to the available inventory of non-nuclear test facilities and is uniquely capable of investigating and characterizing candidate fuel/structural materials, improving associated processing/fabrication techniques, and simulating reactor thermal hydraulics. This paper summarizes facility design and engineering development efforts and reports baseline operational characteristics as determined from a series of performance mapping and long duration capability demonstration tests. Potential follow-on developmental strategies are also suggested in view of the technical and policy challenges ahead. Keywords: Nuclear Rocket Engine, Reactor Environments, Non-Nuclear Testing, Fissile Fuel Development.

Closed Brayton cycle power conversion systems for nuclear reactors :

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wright, Steven A.; Lipinski, Ronald J.; Vernon, Milton E.

2006-04-01

This report describes the results of a Sandia National Laboratories internally funded research program to study the coupling of nuclear reactors to gas dynamic Brayton power conversion systems. The research focused on developing integrated dynamic system models, fabricating a 10-30 kWe closed loop Brayton cycle, and validating these models by operating the Brayton test-loop. The work tasks were performed in three major areas. First, the system equations and dynamic models for reactors and Closed Brayton Cycle (CBC) systems were developed and implemented in SIMULINKTM. Within this effort, both steady state and dynamic system models for all the components (turbines, compressors,more » reactors, ducting, alternators, heat exchangers, and space based radiators) were developed and assembled into complete systems for gas cooled reactors, liquid metal reactors, and electrically heated simulators. Various control modules that use proportional-integral-differential (PID) feedback loops for the reactor and the power-conversion shaft speed were also developed and implemented. The simulation code is called RPCSIM (Reactor Power and Control Simulator). In the second task an open cycle commercially available Capstone C30 micro-turbine power generator was modified to provide a small inexpensive closed Brayton cycle test loop called the Sandia Brayton test-Loop (SBL-30). The Capstone gas-turbine unit housing was modified to permit the attachment of an electrical heater and a water cooled chiller to form a closed loop. The Capstone turbine, compressor, and alternator were used without modification. The Capstone systems nominal operating point is 1150 K turbine inlet temperature at 96,000 rpm. The annular recuperator and portions of the Capstone control system (inverter) and starter system also were reused. The rotational speed of the turbo-machinery is controlled by adjusting the alternator load by using the electrical grid as the load bank. The SBL-30 test loop was operated at the manufacturers site (Barber-Nichols Inc.) and installed and operated at Sandia. A sufficiently detailed description of the loop is provided in this report along with the design characteristics of the turbo-alternator-compressor set to allow other researchers to compare their results with those measured in the Sandia test-loop. The third task consisted of a validation effort. In this task the test loop was operated and compared with the modeled results to develop a more complete understanding of this electrically heated closed power generation system and to validate the model. The measured and predicted system temperatures and pressures are in good agreement, indicating that the model is a reasonable representation of the test loop. Typical deviations between the model and the hardware results are less than 10%. Additional tests were performed to assess the capability of the Brayton engine to continue to remove decay heat after the reactor/heater is shutdown, to develop safe and effective control strategies, and to access the effectiveness of gas inventory control as an alternative means to provide load following. In one test the heater power was turned off to simulate a rapid reactor shutdown, and the turbomachinery was driven solely by the sensible heat stored in the heater for over 71 minutes without external power input. This is an important safety feature for CBC systems as it means that the closed Brayton loop will keep cooling the reactor without the need for auxiliary power (other than that needed to circulate the waste heat rejection coolant) provided the heat sink is available.« less

U.S. Nuclear Cooperation with India: Issues for Congress

DTIC Science & Technology

2008-10-17

safeguards-irrelevant.” The following facilities and activities were not on the separation list: ! 8 indigenous Indian power reactors ! Fast Breeder ...test Reactor (FTBR) and Prototype Fast Breeder Reactors (PFBR) under construction ! Enrichment facilities ! Spent fuel reprocessing facilities (except...potential use in a bomb. In addition, safeguards on enrichment, reprocessing plants, and breeder reactors would support the 2002 U.S. National Strategy to

Direct Production of Propene from the Thermolysis of Poly(..beta..-hydroxybutyrate)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mittal, Ashutosh; Pilath, Heidi M.; Johnson, David K.

To transform biomass components into hydrocarbon fuels it is clear that there are two main transformations that need to occur, i.e., deoxygenation and carbon chain extension. The potential routes for decreasing the oxygen content of biomass intermediates include dehydration, hydrodeoxygenation and decarboxylation. One route that is examined here is the conversion of polyhydroxyalkanoates (PHA) to alkenes that would be intermediates to hydrocarbon fuels.Thermal breakdown of PHA proceeds via an intermediate carboxylic acid, which can then be decarboxylated to an alkene. Oligomerization of alkenes by well-known commercial technologies would permit production of a range of hydrocarbon fuels from a carbohydrate derivedmore » intermediate. Moreover, polyhydroxybutyrate (PHB) can be produced in Cupriavidus necator (formerly known as Ralstonia eutropha) and Alcaligenes eutrophus on a variety of carbon sources including glucose, fructose and glycerol with PHB accumulation reaching 75 percent of dry cell mass. We conducted thermal conversion of PHB and pure crotonic acid (CA), the intermediate carboxylic acid produced by thermal depolymerization of PHB, in a flow-through reactor. The results of initial experiments on the thermal conversion of CA showed that up to 75 mole percent yields of propene could be achieved by optimizing the residence time and temperature of the reactor. Further experiments are being investigated to optimize the reactor parameters and enhance propene yields via thermal conversion of PHB.« less

324 Building B-Cell Pressurized Water Reactor Spent Fuel Packaging & Shipment RL Readiness Assessment Final Report [SEC 1 Thru 3

DOE Office of Scientific and Technical Information (OSTI.GOV)

HUMPHREYS, D C

A parallel readiness assessment (RA) was conducted by independent Fluor Hanford (FH) and U. S. Department of Energy, Richland Operations Office (RL) team to verify that an adequate state of readiness had been achieved for activities associated with the packaging and shipping of pressurized water reactor fuel assemblies from B-Cell in the 324 Building to the interim storage area at the Canister Storage Building in the 200 Area. The RL review was conducted in parallel with the FH review in accordance with the Joint RL/FH Implementation Plan (Appendix B). The RL RA Team members were assigned a FH RA Teammore » counterpart for the review. With this one-on-one approach, the RL RA Team was able to assess the FH Team's performance, competence, and adherence to the implementation plan and evaluate the level of facility readiness. The RL RA Team agrees with the FH determination that startup of the 324 Building B-Cell pressurized water reactor spent nuclear fuel packaging and shipping operations can safely proceed, pending completion of the identified pre-start items in the FH final report (see Appendix A), completion of the manageable list of open items included in the facility's declaration of readiness, and execution of the startup plan to operations.« less

Evaluation of CFETR as a Fusion Nuclear Science Facility using multiple system codes

NASA Astrophysics Data System (ADS)

Chan, V. S.; Costley, A. E.; Wan, B. N.; Garofalo, A. M.; Leuer, J. A.

2015-02-01

This paper presents the results of a multi-system codes benchmarking study of the recently published China Fusion Engineering Test Reactor (CFETR) pre-conceptual design (Wan et al 2014 IEEE Trans. Plasma Sci. 42 495). Two system codes, General Atomics System Code (GASC) and Tokamak Energy System Code (TESC), using different methodologies to arrive at CFETR performance parameters under the same CFETR constraints show that the correlation between the physics performance and the fusion performance is consistent, and the computed parameters are in good agreement. Optimization of the first wall surface for tritium breeding and the minimization of the machine size are highly compatible. Variations of the plasma currents and profiles lead to changes in the required normalized physics performance, however, they do not significantly affect the optimized size of the machine. GASC and TESC have also been used to explore a lower aspect ratio, larger volume plasma taking advantage of the engineering flexibility in the CFETR design. Assuming the ITER steady-state scenario physics, the larger plasma together with a moderately higher BT and Ip can result in a high gain Qfus ˜ 12, Pfus ˜ 1 GW machine approaching DEMO-like performance. It is concluded that the CFETR baseline mode can meet the minimum goal of the Fusion Nuclear Science Facility (FNSF) mission and advanced physics will enable it to address comprehensively the outstanding critical technology gaps on the path to a demonstration reactor (DEMO). Before proceeding with CFETR construction steady-state operation has to be demonstrated, further development is needed to solve the divertor heat load issue, and blankets have to be designed with tritium breeding ratio (TBR) >1 as a target.

Nuclear fuel management optimization using genetic algorithms

DOE Office of Scientific and Technical Information (OSTI.GOV)

DeChaine, M.D.; Feltus, M.A.

1995-07-01

The code independent genetic algorithm reactor optimization (CIGARO) system has been developed to optimize nuclear reactor loading patterns. It uses genetic algorithms (GAs) and a code-independent interface, so any reactor physics code (e.g., CASMO-3/SIMULATE-3) can be used to evaluate the loading patterns. The system is compared to other GA-based loading pattern optimizers. Tests were carried out to maximize the beginning of cycle k{sub eff} for a pressurized water reactor core loading with a penalty function to limit power peaking. The CIGARO system performed well, increasing the k{sub eff} after lowering the peak power. Tests of a prototype parallel evaluation methodmore » showed the potential for a significant speedup.« less

Technology, safety, and costs of decommissioning reference nuclear research and test reactors: sensitivity of decommissioning radiation exposure and costs to selected parameters

DOE Office of Scientific and Technical Information (OSTI.GOV)

Konzek, G.J.

1983-07-01

Additional analyses of decommissioning at the reference research and test (R and T) reactors and analyses of five recent reactor decommissionings are made that examine some parameters not covered in the initial study report (NUREG/CR-1756). The parameters examined for decommissioning are: (1) the effect on costs and radiation exposure of plant size and/or type; (2) the effects on costs of increasing disposal charges and of unavailability of waste disposal capacity at licensed waste disposal facilities; and (3) the costs of and the available alternatives for the disposal of nuclear R and T reactor fuel assemblies.

TEST-HOLE CONSTRUCTION FOR A NEUTRONIC REACTOR

DOEpatents

Ohlinger, L.A.; Seitz, F.; Young, G.J.

1959-02-17

Test-hole construction is described for a reactor which provides safe and ready access to the neutron flux region for specimen materials which are to be irradiated therein. An elongated tubular thimble adapted to be inserted in the access hole through the wall of the reactor is constructed of aluminum and is provided with a plurality of holes parallel to the axis of the thimble for conveying the test specimens into position for irradiation, and a conduit for the circulation of coolant. A laminated shield formed of alternate layers of steel and pressed wood fiber is disposed lengthwise of the thimble near the outer end thereof.

Fabrication of U-10 wt.%Zr Metallic Fuel Rodlets for Irradiation Test in BOR-60 Fast Reactor

DOE PAGES

Kim, Ki-Hwan; Kim, Jong-Hwan; Oh, Seok-Jin; ...

2016-01-01

The fabrication technology for metallic fuel has been developed to produce the driver fuel in a PGSFR in Korea since 2007. In order to evaluate the irradiation integrity and validate the in-reactor of the starting metallic fuel with FMS cladding for the loading of the metallic fuel, U-10 wt.%Zr fuel rodlets were fabricated and evaluated for a verification of the starting driver fuel through an irradiation test in the BOR-60 fast reactor. The injection casting method was applied to U-10 wt.%Zr fuel slugs with a diameter of 5.5 mm. Consequently, fuel slugs per melting batch without casting defects were fabricated through the developmentmore » of advanced casting technology and evaluation tests. The optimal GTAW welding conditions were also established through a number of experiments. In addition, a qualification test was carried out to prove the weld quality of the end plug welding of the metallic fuel rodlets. The wire wrapping of metallic fuel rodlets was successfully accomplished for the irradiation test. Thus, PGSFR fuel rodlets have been soundly fabricated for the irradiation test in a BOR-60 fast reactor.« less

PIE on Safety-Tested Loose Particles from Irradiated Compact 4-4-2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hunn, John D.; Gerczak, Tyler J.; Morris, Robert Noel

2016-04-01

Post-irradiation examination (PIE) is being performed in support of tristructural isotropic (TRISO) coated particle fuel development and qualification for High Temperature Gas-cooled Reactors (HTGRs). This work is sponsored by the Department of Energy Office of Nuclear Energy (DOE-NE) through the Advanced Reactor Technologies (ART) Office under the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program. The AGR-1 experiment was the first in a series of TRISO fuel irradiation tests initiated in 2006. The AGR-1 TRISO particles and fuel compacts were fabricated at Oak Ridge National Laboratory (ORNL) in 2006 using laboratory-scale equipment and irradiated for 3 years in themore » Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to demonstrate and evaluate fuel performance under HTGR irradiation conditions. Post-irradiation examination was performed at INL and ORNL to study how the fuel behaved during irradiation, and to test fuel performance during exposure to elevated temperatures at or above temperatures that could occur during a depressurized conduction cooldown event. This report summarizes safety testing and post-safety testing PIE conducted at ORNL on loose particles extracted from irradiated AGR-1 Compact 4-4-2.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lisowski, Darius D.; Kraus, Adam R.; Bucknor, Matthew D.

A 1/2 scale test facility has been constructed at Argonne National Laboratory to study the heat removal performance and natural circulation flow patterns in a Reactor Cavity Cooling System (RCCS). Our test facility, the Natural convection Shutdown heat removal Test Facility (NSTF), supports the broader goal of developing an inherently safe and fully passive ex-vessel decay heat removal for advanced reactor designs. The project, initiated in 2010 to support the Advanced Reactor Technologies (ART), Small Modular Reactor (SMR), and Next Generation Nuclear Plant (NGNP) programs, has been conducting experimental operations since early 2014. The following paper provides a summary ofmore » some primary design features of the 26-m tall test facility along with a description of the data acquisition suite that guides our experimental practices. Specifics of the distributed fiber optic temperature measurements will be discussed, which introduces an unparalleled level of data density that has never before been implemented in a large scale natural circulation test facility. Results from our test series will then be presented, which provide insight into the thermal hydraulic behavior at steady-state and transient conditions for varying heat flux levels and exhaust chimney configuration states. (C) 2016 Elsevier B.V. All rights reserved.« less

Toward a Mechanistic Source Term in Advanced Reactors: A Review of Past U.S. SFR Incidents, Experiments, and Analyses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bucknor, Matthew; Brunett, Acacia J.; Grabaskas, David

In 2015, as part of a Regulatory Technology Development Plan (RTDP) effort for sodium-cooled fast reactors (SFRs), Argonne National Laboratory investigated the current state of knowledge of source term development for a metal-fueled, pool-type SFR. This paper provides a summary of past domestic metal-fueled SFR incidents and experiments and highlights information relevant to source term estimations that were gathered as part of the RTDP effort. The incidents described in this paper include fuel pin failures at the Sodium Reactor Experiment (SRE) facility in July of 1959, the Fermi I meltdown that occurred in October of 1966, and the repeated meltingmore » of a fuel element within an experimental capsule at the Experimental Breeder Reactor II (EBR-II) from November 1967 to May 1968. The experiments described in this paper include the Run-Beyond-Cladding-Breach tests that were performed at EBR-II in 1985 and a series of severe transient overpower tests conducted at the Transient Reactor Test Facility (TREAT) in the mid-1980s.« less

Annual Symposium, Psychology in the Air Force (5th), 8 April - 10 April 1976

DTIC Science & Technology

1976-06-01

approach . Miller, Robert B., Psychological Considerations in the Design of Training Equipment, WADC-TR-54-563, Wright-Patterson AFB, OH, Wright Air... Abnormal and Social Psychology , 1963, 66, 207-216. Berkley, C. S. & Sproule, C. F. Test anxiety and test-unsophistication: The effects, the cures. Public...SYMPOSIUM PROCEEDINGS Proceedings of the 5th Symposium on Psychology in the Air Force. Sponsored by the Department of Behavioral Sciences and

SAFETY AND SECURITY BUILDING, TRA614. ELEVATIONS. SECTIONS. TWO ROOF LEVELS. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

SAFETY AND SECURITY BUILDING, TRA-614. ELEVATIONS. SECTIONS. TWO ROOF LEVELS. BLAW-KNOX 3150-814-2, 3/1950. INL INDEX NO. 531-0614-00-098-100703, REV. 6. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

SAFETY AND SECURITY BUILDING, TRA614. SIMPLIFIED FLOOR LAYOUT AND WEST ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

SAFETY AND SECURITY BUILDING, TRA-614. SIMPLIFIED FLOOR LAYOUT AND WEST ELEVATION. BLAW-KNOX 3150-14-1, 1/1950. INL INDEX NO. 531-0614-00-098-100024, REV. 2. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

PROCESS WATER BUILDING, TRA605. FLASH EVAPORATOR, CONDENSER (PROJECT FROM EVAPORATOR), ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PROCESS WATER BUILDING, TRA-605. FLASH EVAPORATOR, CONDENSER (PROJECT FROM EVAPORATOR), AND STEAM EJECTOR (ALONG REAR WALL). INL NEGATIVE NO. 4377. M.H. Bartz, Photographer, 3/5/1952 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

FAN HOUSE INTERIOR. THREE MOTOR DRIVES FOR POSITIVE DISPLACEMENT BLOWERS ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

FAN HOUSE INTERIOR. THREE MOTOR DRIVES FOR POSITIVE DISPLACEMENT BLOWERS LINE UP ON NORTH WALL. CONCRETE PEDESTALS. CAMERA FACES NORTHEAST. INL NEGATIVE NO. 4291. Unknown Photographer, 2/26/1952 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

MTR BUILDING AND BALCONY FLOORS. CAMERA FACING EASTERLY. PHOTOGRAPHER DID ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

MTR BUILDING AND BALCONY FLOORS. CAMERA FACING EASTERLY. PHOTOGRAPHER DID NOT EXPLAIN DARK CLOUD. MTR WING WILL ATTACH TO GROUND FLOOR. INL NEGATIVE NO. 1567. Unknown Photographer, 2/28/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR BUILDING, TRA642, INTERIOR. BASEMENT. LIQUID SODIUM PIPING INSIDE CUBICLE ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR BUILDING, TRA-642, INTERIOR. BASEMENT. LIQUID SODIUM PIPING INSIDE CUBICLE SHOWN IN ID-33-G-101. INL NEGATIVE NO. HD24-3-4. Mike Crane, Photographer, 11/2000 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

The development of a universal diagnostic probe system for Tokamak fusion test reactor

NASA Technical Reports Server (NTRS)

Mastronardi, R.; Cabral, R.; Manos, D.

1982-01-01

The Tokamak Fusion Test Reactor (TFTR), the largest such facility in the U.S., is discussed with respect to instrumentation in general and mechanisms in particular. The design philosophy and detailed implementation of a universal probe mechanism for TFTR is discussed.

10 CFR 100.1 - Purpose.

Code of Federal Regulations, 2011 CFR

2011-01-01

... COMMISSION (CONTINUED) REACTOR SITE CRITERIA § 100.1 Purpose. (a) The purpose of this part is to establish approval requirements for proposed sites for stationary power and testing reactors subject to part 50 or part 52 of this chapter. (b) There exists a substantial base of knowledge regarding power reactor...

9 CFR 309.14 - Brucellosis-reactor goats.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 9 Animals and Animal Products 2 2014-01-01 2014-01-01 false Brucellosis-reactor goats. 309.14... INSPECTION AND CERTIFICATION ANTE-MORTEM INSPECTION § 309.14 Brucellosis-reactor goats. Goats which have reacted to a test for brucellosis shall not be slaughtered in an official establishment. ...

9 CFR 309.14 - Brucellosis-reactor goats.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 9 Animals and Animal Products 2 2013-01-01 2013-01-01 false Brucellosis-reactor goats. 309.14... INSPECTION AND CERTIFICATION ANTE-MORTEM INSPECTION § 309.14 Brucellosis-reactor goats. Goats which have reacted to a test for brucellosis shall not be slaughtered in an official establishment. ...

9 CFR 309.14 - Brucellosis-reactor goats.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 9 Animals and Animal Products 2 2012-01-01 2012-01-01 false Brucellosis-reactor goats. 309.14... INSPECTION AND CERTIFICATION ANTE-MORTEM INSPECTION § 309.14 Brucellosis-reactor goats. Goats which have reacted to a test for brucellosis shall not be slaughtered in an official establishment. ...

FFTF Passive Safety Test Data for Benchmarks for New LMR Designs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wootan, David W.; Casella, Andrew M.

Liquid Metal Reactors (LMRs) continue to be considered as an attractive concept for advanced reactor design. Software packages such as SASSYS are being used to im-prove new LMR designs and operating characteristics. Significant cost and safety im-provements can be realized in advanced liquid metal reactor designs by emphasizing inherent or passive safety through crediting the beneficial reactivity feedbacks associ-ated with core and structural movement. This passive safety approach was adopted for the Fast Flux Test Facility (FFTF), and an experimental program was conducted to characterize the structural reactivity feedback. The FFTF passive safety testing pro-gram was developed to examine howmore » specific design elements influenced dynamic re-activity feedback in response to a reactivity input and to demonstrate the scalability of reactivity feedback results to reactors of current interest. The U.S. Department of En-ergy, Office of Nuclear Energy Advanced Reactor Technology program is in the pro-cess of preserving, protecting, securing, and placing in electronic format information and data from the FFTF, including the core configurations and data collected during the passive safety tests. Benchmarks based on empirical data gathered during operation of the Fast Flux Test Facility (FFTF) as well as design documents and post-irradiation examination will aid in the validation of these software packages and the models and calculations they produce. Evaluation of these actual test data could provide insight to improve analytical methods which may be used to support future licensing applications for LMRs« less

Advanced Demonstration and Test Reactor Options Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Petti, David Andrew; Hill, R.; Gehin, J.

Global efforts to address climate change will require large-scale decarbonization of energy production in the United States and elsewhere. Nuclear power already provides 20% of electricity production in the United States (U.S.) and is increasing in countries undergoing rapid growth around the world. Because reliable, grid-stabilizing, low emission electricity generation, energy security, and energy resource diversity will be increasingly valued, nuclear power’s share of electricity production has a potential to grow. In addition, there are non electricity applications (e.g., process heat, desalination, hydrogen production) that could be better served by advanced nuclear systems. Thus, the timely development, demonstration, and commercializationmore » of advanced nuclear reactors could diversify the nuclear technologies available and offer attractive technology options to expand the impact of nuclear energy for electricity generation and non-electricity missions. The purpose of this planning study is to provide transparent and defensible technology options for a test and/or demonstration reactor(s) to be built to support public policy, innovation and long term commercialization within the context of the Department of Energy’s (DOE’s) broader commitment to pursuing an “all of the above” clean energy strategy and associated time lines. This planning study includes identification of the key features and timing needed for advanced test or demonstration reactors to support research, development, and technology demonstration leading to the commercialization of power plants built upon these advanced reactor platforms. This planning study is consistent with the Congressional language contained within the fiscal year 2015 appropriation that directed the DOE to conduct a planning study to evaluate “advanced reactor technology options, capabilities, and requirements within the context of national needs and public policy to support innovation in nuclear energy”. Advanced reactors are defined in this study as reactors that use coolants other than water. Advanced reactor technologies have the potential to expand the energy applications, enhance the competitiveness, and improve the sustainability of nuclear energy.« less

A microprocessor tester for the treat upgrade reactor trip system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lenkszus, F.R.; Bucher, R.G.

1985-02-01

The upgrading of the Transient Reactor Test (TREAT) Facility at ANL-Idaho has been designed to provide additional experimental capabilities for the study of core disruptive accident (CDA) phenomena. To improve the analytical extrapolation of test results to full-size assembly bundles, the facility upgrade will increase the maximum size of the test bundle from 7 to 37 fuel pins. By creating a core convertor zone around the test location, the neutron spectrum incident on the test assembly will be hardened and the maximum energy deposited in the sample will be increased. In addition, a programmable Automated Reactor Control System (ARCS) willmore » permit high-power transients up to 11,000 MW having a controlled reactor period of from 15 to 0.1 sec. These modifications to the core neutronics will improve simulation of LMFBR accident conditions. Finally, a sophisticated, multiply-redundant safety system, the Reactor Trip System (RTS), will provide safe operation for both steady state and transient production operating modes. To insure that this complex safety system is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations. A quantitative reliability analysis of the RTS shows that the unreliability, that is, the probability of failure, is acceptable for a 10 hour mission time or risk interval.« less

UO{sub 2} and PuO{sub 2} utilization in high temperature engineering test reactor with helium coolant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Waris, Abdul, E-mail: awaris@fi.itb.ac.id; Novitrian,; Pramuditya, Syeilendra

High temperature engineering test reactor (HTTR) is one of high temperature gas cooled reactor (HTGR) types which has been developed by Japanese Atomic Energy Research Institute (JAERI). The HTTR is a graphite moderator, helium gas coolant, 30 MW thermal output and 950 °C outlet coolant temperature for high temperature test operation. Original HTTR uses UO{sub 2} fuel. In this study, we have evaluated the use of UO{sub 2} and PuO{sub 2} in form of mixed oxide (MOX) fuel in HTTR. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. Themore » result shows that HTTR can obtain its criticality condition if the enrichment of {sup 235}U in loaded fuel is 18.0% or above.« less

Experiment for search for sterile neutrino at SM-3 reactor

NASA Astrophysics Data System (ADS)

Serebrov, A. P.; Ivochkin, V. G.; Samoylov, R. M.; Fomin, A. K.; Zinoviev, V. G.; Neustroev, P. V.; Golovtsov, V. L.; Gruzinsky, N. V.; Solovey, V. A.; Cherniy, A. V.; Zherebtsov, O. M.; Martemyanov, V. P.; Zinoev, V. G.; Tarasenkov, V. G.; Aleshin, V. I.; Petelin, A. L.; Pavlov, S. V.; Izhutov, A. L.; Sazontov, S. A.; Ryazanov, D. K.; Gromov, M. O.; Afanasiev, V. V.; Matrosov, L. N.; Matrosova, M. Yu.

2016-11-01

In connection with the question of possible existence of sterile neutrino the laboratory on the basis of SM-3 reactor was created to search for oscillations of reactor antineutrino. A prototype of a neutrino detector with scintillator volume of 400 l can be moved at the distance of 6-11 m from the reactor core. The measurements of background conditions have been made. It is shown that the main experimental problem is associated with cosmic radiation background. Test measurements of dependence of a reactor antineutrino flux on the distance from a reactor core have been made. The prospects of search for oscillations of reactor antineutrino at short distances are discussed.

Zirconium Hydride Space Power Reactor design.

NASA Technical Reports Server (NTRS)

Asquith, J. G.; Mason, D. G.; Stamp, S.

1972-01-01

The Zirconium Hydride Space Power Reactor being designed and fabricated at Atomics International is intended for a wide range of potential applications. Throughout the program a series of reactor designs have been evaluated to establish the unique requirements imposed by coupling with various power conversion systems and for specific applications. Current design and development emphasis is upon a 100 kilowatt thermal reactor for application in a 5 kwe thermoelectric space power generating system, which is scheduled to be fabricated and ground tested in the mid 70s. The reactor design considerations reviewed in this paper will be discussed in the context of this 100 kwt reactor and a 300 kwt reactor previously designed for larger power demand applications.

Evaluating and planning the radioactive waste options for dismantling the Tokamak Fusion Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rule, K.; Scott, J.; Larson, S.

1995-12-31

The Tokamak Fusion Test Reactor (TFTR) is a one-of-a kind tritium fusion research reactor, and is planned to be decommissioned within the next several years. This is the largest fusion reactor in the world and as a result of deuterium-tritum reactions is tritium contaminated and activated from 14 Mev neutrons. This presents many unusual challenges when dismantling, packaging and disposing its components and ancillary systems. Special containers are being designed to accommodate the vacuum vessel, neutral beams, and tritium delivery and processing systems. A team of experienced professionals performed a detailed field study to evaluate the requirements and appropriate methodsmore » for packaging the radioactive materials. This team focused on several current and innovative methods for waste minimization that provides the oppurtunmost cost effective manner to package and dispose of the waste. This study also produces a functional time-phased schedule which conjoins the waste volume, weight, costs and container requirements with the detailed project activity schedule for the entire project scope. This study and project will be the first demonstration of the decommissioning of a tritium fusion test reactor. The radioactive waste disposal aspects of this project are instrumental in demonstrating the viability of a fusion power reactor with regard to its environmental impact and ultimate success.« less

Application of a fluidized bed reactor charged with aragonite for control of alkalinity, pH and carbon dioxide in marine recirculating aquaculture systems

USGS Publications Warehouse

Paul S Wills, PhD; Pfeiffer, Timothy; Baptiste, Richard; Watten, Barnaby J.

2016-01-01

Control of alkalinity, dissolved carbon dioxide (dCO2), and pH are critical in marine recirculating aquaculture systems (RAS) in order to maintain health and maximize growth. A small-scale prototype aragonite sand filled fluidized bed reactor was tested under varying conditions of alkalinity and dCO2 to develop and model the response of dCO2 across the reactor. A large-scale reactor was then incorporated into an operating marine recirculating aquaculture system to observe the reactor as the system moved toward equilibrium. The relationship between alkalinity dCO2, and pH across the reactor are described by multiple regression equations. The change in dCO2 across the small-scale reactor indicated a strong likelihood that an equilibrium alkalinity would be maintained by using a fluidized bed aragonite reactor. The large-scale reactor verified this observation and established equilibrium at an alkalinity of approximately 135 mg/L as CaCO3, dCO2 of 9 mg/L, and a pH of 7.0 within 4 days that was stable during a 14 day test period. The fluidized bed aragonite reactor has the potential to simplify alkalinity and pH control, and aid in dCO2 control in RAS design and operation. Aragonite sand, purchased in bulk, is less expensive than sodium bicarbonate and could reduce overall operating production costs.

Initial Back-to-Back Fission Chamber Testing in ATRC

DOE Office of Scientific and Technical Information (OSTI.GOV)

Benjamin Chase; Troy Unruh; Joy Rempe

2014-06-01

Development and testing of in-pile, real-time neutron sensors for use in Materials Test Reactor experiments is an ongoing project at Idaho National Laboratory. The Advanced Test Reactor National Scientific User Facility has sponsored a series of projects to evaluate neutron detector options in the Advanced Test Reactor Critical Facility (ATRC). Special hardware was designed and fabricated to enable testing of the detectors in the ATRC. Initial testing of Self-Powered Neutron Detectors and miniature fission chambers produced promising results. Follow-on testing required more experiment hardware to be developed. The follow-on testing used a Back-to-Back fission chamber with the intent to providemore » calibration data, and a means of measuring spectral indices. As indicated within this document, this is the first time in decades that BTB fission chambers have been used in INL facilities. Results from these fission chamber measurements provide a baseline reference for future measurements with Back-to-Back fission chambers.« less

Lunar Regolith Simulant Feed System for a Hydrogen Reduction Reactor System

NASA Technical Reports Server (NTRS)

Mueller, R. P.; Townsend, Ivan I., III

2009-01-01

One of the goals of In-Situ Resource Utilization (ISRU) on the moon is to produce oxygen from the lunar regolith which is present in the form of Ilmenite (FeTi03) and other compounds. A reliable and attainable method of extracting some of the oxygen from the lunar regolith is to use the hydrogen reduction process in a hot reactor to create water vapor which is then condensed and electrolyzed to obtain oxygen for use as a consumable. One challenge for a production system is to reliably acquire the regolith with an excavator hauler mobility platform and then introduce it into the reactor inlet tube which is raised from the surface and above the reactor itself. After the reaction, the hot regolith (-1000 C) must be expelled from the reactor for disposal by the excavator hauler mobility system. In addition, the reactor regolith inlet and outlet tubes must be sealed by valves during the reaction in order to allow collection of the water vapor by the chemical processing sub-system. These valves must be able to handle abrasive regolith passing through them as well as the heat conduction from the hot reactor. In 2008, NASA has designed and field tested a hydrogen reduction system called ROxygen in order to demonstrate the feasibility of extracting oxygen from lunar regolith. The field test was performed with volcanic ash known as Tephra on Mauna Kea volcano on the Big Island of Hawai'i. The tephra has similar properties to lunar regolith, so that it is regarded as a good simulant for the hydrogen reduction process. This paper will discuss the design, fabrication, operation, test results and lessons learned with the ROxygen regolith feed system as tested on Mauna Kea in November 2008.

Analysis and Down Select of Flow Passages for Thermal Hydraulic Testing of a SNAP Derived Reactor

NASA Technical Reports Server (NTRS)

Godfroy, T. J.; Sadasivan, P.; Masterson, S.

2007-01-01

As past of the Vision for Space Exploration, man will return to the moon. To enable safe and productive time on the lunar surface will require adequate power resources. To provide the needed power and to give mission planners all landing site possibilities, including a permanently dark crater, a nuclear reactor provides the most options. Designed to be l00kWt providing approx. 25kWe this power plants would be very effective in delivering dependable, site non-specific power to crews or robotic missions on the lunar surface. An affordable reference reactor based upon the successful SNAP program of the 1960's and early 1970's has been designed by Los Alamos National Laboratory that will meet such a requirement. Considering current funding, environmental, and schedule limitations this lunar surface power reactor will be tested using non-nuclear simulators to simulate the heat from fission reactions. Currently a 25kWe surface power SNAP derivative reactor is in the early process of design and testing with collaboration between Los Alamos National Laboratory, Idaho National Laboratory, Glenn Research Center, Marshall Space Flight Center, and Sandia National Laboratory to ensure that this new design is affordable and can be tested using non-nuclear methods as have proven so effective in the past. This paper will discuss the study and down selection of a flow passage concept for a approx. 25kWe lunar surface power reactor. Several different flow passages designs were evaluated using computational fluid dynamics to determine pressure drop and a structural assessment to consider thermal and stress of the passage walls. The reactor design basis conditions are discussed followed by passage problem setup and results for each concept. A recommendation for passage design is made with rationale for selection.

The high-temperature oxidation of aromatic hydrocarbons

NASA Technical Reports Server (NTRS)

Brezinsky, K.

1986-01-01

Chemical mechanisms of the atmospheric pressure, high-temperature (875-1500 K) gas-phase oxidation of benzene, toluene, ethylbenzene, and propylbenzene are described and discussed. Oxidation trends evident from turbulent flow reactor experiments serve as the basis for the mechanisms of the oxidation of benzene and alkylated aromatics. The potential effects of very high temperatures and pressures on the chemistry of oxidation of aromatics are described. The oxidation of benzene and phenyl radical has been found to proceed in a stepwise C6-C5-C4 sequence. Species profiles obtained from flow-reactor experiments suggest that the oxidation of benzene and phenyl radical follows the generalized route via phenoxy, cyclopentadienyl and butadienyl radical. The oxidation of the C4 species branches into multiple pathways that yield copious amounts of ethylene and acetylene. Certain major trends are evident: the alkylated aromatics on initial attack either form styrene, benzyl radical or benzene. The styrene reacts further to produce a benzyl radical or benzene. The oxidation of an alkylated aromatic hydrocarbon appears eventually to reduce to the oxidation of either phenyl radical or benzene.

1985 Nuclear Science Symposium, 32nd, and 1985 Symposium on Nuclear Power Systems, 17th, San Francisco, CA, October 23-25, 1985, Proceedings

NASA Technical Reports Server (NTRS)

1986-01-01

The present conference ranges over topics in high energy physics instrumentation, detectors, nuclear medical applications, health physics and environmental monitoring, reactor instrumentation, nuclear spacecraft instrumentation, the 'Fastbus' data acquisition system, circuits and systems for nuclear research facilities, and the development status of nuclear power systems. Specific attention is given to CCD high precision detectors, a drift chamber preamplifier, a Cerenkov ring imaging detector, novel scintillation glasses and scintillating fibers, a modular multidrift vertex detector, radial wire drift chambers, liquid argon polarimeters, a multianode photomultiplier, the reliability of planar silicon detectors, the design and manufacture of wedge and strip anodes, ultrafast triode photodetectors, photomultiplier tubes, a barium fluoride plastic scintillator, a fine grained neutron hodoscope, the stability of low leakage silicon photodiodes for crystal calorimeters, and X-ray proportional counters. Also considered are positron emission tomography, single photon emission computed tomography, nuclear magnetic resonance imaging, Geiger-Muller detectors, nuclear plant safeguards, a 32-bit Fastbus computer, an advanced light water reactor, and nuclear plant maintenance.

Heat Pipe Reactor Dynamic Response Tests: SAFE-100 Reactor Core Prototype

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.

2005-01-01

The SAFE-I00a test article at the NASA Marshall Space Flight Center was used to simulate a variety of potential reactor transients; the SAFEl00a is a resistively heated, stainless-steel heat-pipe (HP)-reactor core segment, coupled to a gas-flow heat exchanger (HX). For these transients the core power was controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. This type of non-nuclear test is expected to provide reasonable approximation of reactor transient behavior because reactivity feedback is very simple in a compact fast reactor (simple, negative, and relatively monotonic temperature feedback, caused mostly by thermal expansion) and calculations show there are no significant reactivity effects associated with fluid in the HP (the worth of the entire inventory of Na in the core is .

Spacecraft Charging Technology, 1980

NASA Technical Reports Server (NTRS)

1981-01-01

The third Spacecraft Charging Technology Conference proceedings contain 66 papers on the geosynchronous plasma environment, spacecraft modeling, charged particle environment interactions with spacecraft, spacecraft materials characterization, and satellite design and testing. The proceedings is a compilation of the state of the art of spacecraft charging and environmental interaction phenomena.

Risk factors for bovine tuberculosis in low incidence regions related to the movements of cattle

PubMed Central

2013-01-01

Background Bovine tuberculosis (bTB) remains difficult to eradicate from low incidence regions partly due to the imperfect sensitivity and specificity of routine intradermal tuberculin testing. Herds with unconfirmed reactors that are incorrectly classified as bTB-negative may be at risk of spreading disease, while those that are incorrectly classified as bTB-positive may be subject to costly disease eradication measures. This analysis used data from Scotland in the period leading to Officially Tuberculosis Free recognition (1) to investigate the risks associated with the movements of cattle from herds with different bTB risk classifications and (2) to identify herd demographic characteristics that may aid in the interpretation of tuberculin testing results. Results From 2002 to 2009, for every herd with confirmed bTB positive cattle identified through routine herd testing, there was an average of 2.8 herds with at least one unconfirmed positive reactor and 18.9 herds with unconfirmed inconclusive reactors. Approximately 75% of confirmed bTB positive herds were detected through cattle with no known movements outside Scotland. At the animal level, cattle that were purchased from Scottish herds with unconfirmed positive reactors and a recent history importing cattle from endemic bTB regions were significantly more likely to react positively on routine intradermal tuberculin tests, while cattle purchased from Scottish herds with unconfirmed inconclusive reactors were significantly more likely to react inconclusively. Case-case comparisons revealed few demographic differences between herds with confirmed positive, unconfirmed positive, and unconfirmed inconclusive reactors, which highlights the difficulty in determining the true disease status of herds with unconfirmed tuberculin reactors. Overall, the risk of identifying reactors through routine surveillance decreased significantly over time, which may be partly attributable to changes in movement testing regulations and the volume of cattle imported from endemic regions. Conclusions Although the most likely source of bTB infections in Scotland was cattle previously imported from endemic regions, we found indirect evidence of transmission within Scottish cattle farms and cannot rule out the possibility of low level transmission between farms. Further investigation is needed to determine whether targeting herds with unconfirmed reactors and a history of importing cattle from high risk regions would benefit control efforts. PMID:24206865

REVIEW OF POWER AND HEAT REACTOR DESIGNS. Domestic and Foreign

DOE Office of Scientific and Technical Information (OSTI.GOV)

Appleby, E.R., comp

1963-10-01

Unclassified information from domestic and foreign literature from January 1952 through September 1963 is compiled. Design characteristics and current information on the status of the individual designs are given, along with references for the associated literature. SNAP systems, proposed reactors, and chemonuclear and test reactors with characteristics similar to power reactors are included. The designs are indexed by name, location, type, and some special characteristics. (D.C.W.)

Small reactor power system for space application

NASA Technical Reports Server (NTRS)

Shirbacheh, M.

1987-01-01

A development history and comparative performance capability evaluation is presented for spacecraft nuclear powerplant Small Reactor Power System alternatives. The choice of power conversion technology depends on the reactor's operating temperature; thermionic, thermoelectric, organic Rankine, and Alkali metal thermoelectric conversion are the primary power conversion subsystem technology alternatives. A tabulation is presented for such spacecraft nuclear reactor test histories as those of SNAP-10A, SP-100, and NERVA.

Susceptibility constants of airborne bacteria to dielectric barrier discharge for antibacterial performance evaluation.

PubMed

Park, Chul Woo; Hwang, Jungho

2013-01-15

Dielectric barrier discharge (DBD) is a promising method to remove contaminant bioaerosols. The collection efficiency of a DBD reactor is an important factor for determining a reactor's removal efficiency. Without considering collection, simply defining the inactivation efficiency based on colony counting numbers for DBD as on and off may lead to overestimation of the inactivation efficiency of the DBD reactor. One-pass removal tests of bioaerosols were carried out to deduce the inactivation efficiency of the DBD reactor using both aerosol- and colony-counting methods. Our DBD reactor showed good performance for removing test bioaerosols for an applied voltage of 7.5 kV and a residence time of 0.24s, with η(CFU), η(Number), and η(Inactivation) values of 94%, 64%, and 83%, respectively. Additionally, we introduce the susceptibility constant of bioaerosols to DBD as a quantitative parameter for the performance evaluation of a DBD reactor. The modified susceptibility constant, which is the ratio of the susceptibility constant to the volume of the plasma reactor, has been successfully demonstrated for the performance evaluation of different sized DBD reactors under different DBD operating conditions. Our methodology will be used for design optimization, performance evaluation, and prediction of power consumption of DBD for industrial applications. Copyright © 2012 Elsevier B.V. All rights reserved.

10 CFR 830 Major Modification Determination for the Advanced Test Reactor Remote Monitoring and Management Capability

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bohachek, Randolph Charles

2015-09-01

The Advanced Test Reactor (ATR; TRA-670), which is located in the ATR Complex at Idaho National Laboratory, was constructed in the 1960s for the purpose of irradiating reactor fuels and materials. Other irradiation services, such as radioisotope production, are also performed at ATR. While ATR is safely fulfilling current mission requirements, assessments are continuing. These assessments intend to identify areas to provide defense–in-depth and improve safety for ATR. One of the assessments performed by an independent group of nuclear industry experts recommended that a remote accident management capability be provided. The report stated that: “contemporary practice in commercial power reactorsmore » is to provide a remote shutdown station or stations to allow shutdown of the reactor and management of long-term cooling of the reactor (i.e., management of reactivity, inventory, and cooling) should the main control room be disabled (e.g., due to a fire in the control room or affecting the control room).” This project will install remote reactor monitoring and management capabilities for ATR. Remote capabilities will allow for post scram reactor management and monitoring in the event the main Reactor Control Room (RCR) must be evacuated.« less

An underground nuclear power station using self-regulating heat-pipe controlled reactors

DOEpatents

Hampel, V.E.

1988-05-17

A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working fluid in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast- acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor. 5 figs.

Underground nuclear power station using self-regulating heat-pipe controlled reactors

DOEpatents

Hampel, Viktor E.

1989-01-01

A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working flud in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast-acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Boyack, B.E.

The PIUS reactor utilizes simplified, inherent, passive, or other innovative means to accomplish safety functions. Accordingly, the PIUS reactor is subject to the requirements of 10CFR52.47(b)(2)(i)(A). This regulation requires that the applicant adequately demonstrate the performance of each safety feature, interdependent effects among the safety features, and a sufficient data base on the safety features of the design to assess the analytical tools used for safety analysis. Los Alamos has assessed the quality and completeness of the existing and planned data bases used by Asea Brown Boveri to validate its safety analysis codes and other relevant data bases. Only amore » limited data base of separate effect and integral tests exist at present. This data base is not adequate to fulfill the requirements of 10CFR52.47(b)(2)(i)(A). Asea Brown Boveri has stated that it plans to conduct more separate effect and integral test programs. If appropriately designed and conducted, these test programs have the potential to satisfy most of the data base requirements of 10CFR52.47(b)(2)(i)(A) and remedy most of the deficiencies of the currently existing combined data base. However, the most important physical processes in PIUS are related to reactor shutdown because the PIUS reactor does not contain rodded shutdown and control systems. For safety-related reactor shutdown, PIUS relies on negative reactivity insertions from the moderator temperature coefficient and from boron entering the core from the reactor pool. Asea Brown Boveri has neither developed a direct experimental data base for these important processes nor provided a rationale for indirect testing of these key PIUS processes. This is assessed as a significant shortcoming. In preparing the conclusions of this report, test documentation and results have been reviewed for only one integral test program, the small-scale integral tests conducted in the ATLE facility.« less

Pneumatic Regolith Transfer Systems for In Situ Resource Utilization

NASA Technical Reports Server (NTRS)

Mueller, R. P.; Townsend, I. I.; Mantovani, J. G.; Zacny, Kris A.; Craft, Jack

2010-01-01

This slide presentation reviews the testing of a pneumatic system for transfering regolith, to be used for In Situ Resource Utilization (ISRU). Using both the simulated microgravity of parabolic flight and ground testing, the tests demonstrated that lunar regolith can be conveyed pneumatically into a simulated ISRU oxygen production plant reactor. The ground testing also demonstrated that the regolith can be expelled from the ISRU reactor for disposal or for other resource processing.

ETR ELECTRICAL BUILDING, TRA648. EMERGENCY STANDBY GENERATOR AND DIESEL UNIT. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR ELECTRICAL BUILDING, TRA-648. EMERGENCY STANDBY GENERATOR AND DIESEL UNIT. METAL ROOF AND PUMICE BLOCK WALLS. CAMERA FACING SOUTHWEST. INL NEGATIVE NO. 56-3708. R.G. Larsen, Photographer, 11/13/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR ELECTRICAL BUILDING, TRA648. FLOOR PLANS FOR FIRST FLOOR AND ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR ELECTRICAL BUILDING, TRA-648. FLOOR PLANS FOR FIRST FLOOR AND BASEMENT. SECTIONS. KAISER ETR-5528-MTR-648-A-2, 12/1955. INL INDEX NO. 532-0648-00-486-101402, REV. 6. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR BUILDING, TRA642, INTERIOR. CONSOLE FLOOR, SOUTH HALF. CABLE TUNNEL. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR BUILDING, TRA-642, INTERIOR. CONSOLE FLOOR, SOUTH HALF. CABLE TUNNEL. CAMERA FACING SOUTH INTO ETR ELECTRICAL BUILDING (TRA-648). INL NEGATIVE NO. HD46-20-2. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR ELECTRICAL BUILDING, TRA648. ELEVATIONS AND DETAILS. ROOF PLAN. DOOR ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR ELECTRICAL BUILDING, TRA-648. ELEVATIONS AND DETAILS. ROOF PLAN. DOOR SCHEDULE. KAISER ETR-5528-MTR-648-A-3, 1/1956. INL INDEX NO. 532-0648-00-486-101403, REV. 5. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR BUILDING, TRA642, INTERIOR. BASEMENT. ROLLUP DOOR TO CUBICLE POSTS ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR BUILDING, TRA-642, INTERIOR. BASEMENT. ROLLUP DOOR TO CUBICLE POSTS CAUTION SIGNS BECAUSE OF SODIUM HAZARD WITHIN. INL NEGATIVE NO. HD24-3-1. Mike Crane, Photographer, 11/2000 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

78 FR 53482 - Entergy Operations, Inc., River Bend Station, Unit 1; Exemption

Federal Register 2010, 2011, 2012, 2013, 2014

2013-08-29

... facility consists of a boiling-water reactor located in West Feliciana Parish, Louisiana. 2.0 Request... Containment Leakage Testing for Water- Cooled Power Reactors,'' requires that components which penetrate containment be periodically leak tested at the ``P a, '' defined as the ``calculated peak containment internal...

MTR WING, TRA604, INTERIOR. BASEMENT. WEST CORRIDOR. CAMERA FACES NORTH. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

MTR WING, TRA-604, INTERIOR. BASEMENT. WEST CORRIDOR. CAMERA FACES NORTH. HVAC AREA IS AT RIGHT OF CORRIDOR. INL NEGATIVE NO. HD46-13-3. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR HEAT EXCHANGER BUILDING, TRA644. DETAIL OF SOUTH SIDE BUILDING ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR HEAT EXCHANGER BUILDING, TRA-644. DETAIL OF SOUTH SIDE BUILDING INSET. DEMINERALIZER WING AT RIGHT. CAMERA FACING NORTH. INL NEGATIVE NO. HD46-36-2. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID